Lymphatic involvement in the histopathogenesis of mucous retention cyst.

Science.gov (United States)

Kundu, Sukalyan; Cheng, Jun; Maruyama, Satoshi; Suzuki, Makoto; Kawashima, Hiroyuki; Saku, Takashi

2007-01-01

Mucous retention cyst results from extravasation of saliva. Our intent was to study the role of lymphatics in its pathogenesis. Twenty-three surgical specimens of mucous retention cyst of the lip were examined for involvement of lymphatic vessels by a comparative immunohistochemical demonstration of lymphatic and blood vascular endothelial cells, as well as lymphatic and salivary contents. Mucous retention cysts were histopathologically classified into three stages: early, intermediate, and advanced. In the early stage, there was diffuse extravasation of mucous material in the interstitium of the lamina propria or the submucosal layer of the oral mucosa. In the intermediate stage, lymphatics, which were clearly revealed and immunohistochemically distinguished from blood vessels by monoclonal antibody D2-40, were dilated and finally ruptured, leaving fragments of lymphatic walls in the periphery of mucous pools. In the advanced stage, thick cyst walls of granulation tissue were formed around mucous retention. Lymphatics were no longer involved in the granulation tissue wall, which was actively driven by blood vessel formation. The results suggest that the lymphatic rupture seems to contribute to the enlargement in the pathogenesis of mucous retention cyst.

OECD/CSNI Workshop on In-Vessel Core Debris Retention and Coolability - Summary and Conclusions

International Nuclear Information System (INIS)

Behbahani, Ali-Reza; Drozd, Andrzej; Kim, Sang-Baik; Micaelli, Jean-Claude; Okkonen, Timo; Sugimoto, Jun; Trambauer, Klaus; Tuomisto, Harri

1999-01-01

In the spring of 1994 an OECD Workshop on Large Pool Heat transfer was held in Grenoble. The scope of this workshop was the investigation of (1) molten pool heat transfer, (2) heat transfer to the surrounding water, and (3) the feasibility of in-vessel core debris cooling through external cooling of the vessel. Since this time, experimental test series have been completed (e.g., COPO, ULPU, CORVIS) and new experimental programs (e.g., BALI, SONATA, RASPLAV, debris and gap heat transfer) have been established to consolidate and expand the data base for further model development and to improve the understanding of in-vessel debris retention and coolability in a nuclear power plant. Discussions within the CSNI's PWG-2 and the Task Group on Degraded Core Cooling (TG-DCC) have led to the conclusion that the time was ripe for organizing a new international Workshop with the objectives: - to review the results of experimental research that has been conducted in this area; - to exchange information on the results of member countries experiments and model development on in-vessel core debris retention and coolability; - to discuss areas where additional experimental research is needed in order to provide an adequate data base for analytical model development for core debris retention and coolability. The scope of this workshop was limited to the phenomena connected to in-vessel core debris retention and coolability and did not include steam explosion and fission product issues. The workshop was structured into the following sessions: Key note papers; Experiments and model development; Debris bed heat transfer; Corium properties, molten pool convection and crust formation; Gap formation and gap cooling; Creep behaviour of reactor pressure vessel lower head; Ex-vessel boiling and critical heat flux phenomena; Scaling to reactor severe accident conditions and reactor applications. Compared to the previous workshop held in Grenoble in 1994, large progress has been made in the

In vessel retention for VVER 1000 - Experimental work

International Nuclear Information System (INIS)

Batek, D.

2015-01-01

After Fukushima accident, the nuclear community realized that it is necessary to have strategy and solution for severe accident management. In Vessel Retention (IVR) of corium is an important strategy to mitigate the consequences of a severe accident. In this poster the author reviews the present status of experimental works made by UJV (Czech Republic) from 2012 until now, on the IVR strategy specifically applied for the VVER 1000 unit. The BESTH 1 experiment was prepared to test the behavior of the RPV (Reactor Pressure Vessel) surface under 2 configurations: clean and corroded. BESTH 2 experiment is a modification of BESTH 1 experiment in order to get greater thermal fluxes. The BESTH 3 facility is a large scale experiment that is under extensive design (2016-2017) whose main objective will be to investigate the results of vast analytical works made by experts with specialization of severe accident phenomenology

Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

International Nuclear Information System (INIS)

Park, J.W.; Bae, J.H.; Seol, W.C.

2015-01-01

An integrated conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-layered stratified corium (metallic pool, oxide pool and heavy metal) and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel. The 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

In-vessel retention modeling capabilities in MAAP5

International Nuclear Information System (INIS)

Paik, Chan Y.; Lee, Sung Jin; Zhou, Quan; Luangdilok, W.; Reeves, R.W.; Henry, R.E.; Plys, M.; Scobel, J.H.

2012-01-01

Modular Accident Analysis Program (MAAP) is an integrated severe accident analysis code for both light water and heavy water reactors. New and improved models to address the complex phenomena associated with in-vessel retention (IVR) were incorporated into MAAP5.01. They include: -a) time-dependent volatile and non-volatile decay heat, -b) material properties at high temperatures, -c) finer vessel wall nodalization, -d) new correlations for natural convection heat transfer in the oxidic pool, -e) refined metal layer heat transfer to the reactor vessel wall and surroundings, -f) formation of a heavy metal layer, and -g) insulation cooling channel model and associated ex-vessel heat transfer and critical heat flux correlations. In this paper, the new and improved models in MAAP5.01 are described and sample calculation results are presented for the AP1000 passive plant. For the IVR evaluation, a transient calculation is useful because the timing of corium relocation, decaying heat load, and formation of separate layers in the lower plenum all affect integrity of the lower head. The key parameters affecting the IVR success are the metal layer emissivity and thickness of the top metal layer, which depends on the amount of steel in the oxidic pool and in the heavy metal layer. With the best estimate inputs for the debris mixing parameters in a conservative IVR scenario, the AP1000 plant results show that the maximum ex-vessel heat flux to CHF ratio is about 0.7, which occurs before 10.000 seconds when the decay heat is high. The AP1000 plant results demonstrate how MAAP5.01 can be used to evaluate IVR and to gain insight into responses of the lower head during a severe accident

Oxidation effect on steel corrosion and thermal loads during corium melt in-vessel retention

Energy Technology Data Exchange (ETDEWEB)

Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Almjashev, V.I. [Alexandrov Scientific-Research Technology Institute (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [KTH, Stockholm (Sweden); Gusarov, V.V. [SPb State Technology University (SPbGTU), St. Petersburg (Russian Federation); Barrachin, M. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [EC-Joint Research Centre, Institute for Transuranium Elements (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [Commissariat à l’Energie Atomique et aux Energies Alternatives (CEA), Cadarache, St Paul lez Durance (France)

2014-10-15

Highlights: • The METCOR facility simulates vessel steel corrosion in contact with corium. • Steel corrosion rates in UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} coria accelerate above 1050 K. • However corrosion rates can also be limited by melt O{sub 2} supply. • The impact of this on in-vessel retention (IVR) strategy is discussed. - Abstract: During a severe accident with core meltdown, the in-vessel molten core retention is challenged by the vessel steel ablation due to thermal and physicochemical interaction of melt with steel. In accidents with oxidizing atmosphere above the melt surface, a low melting point UO{sub 2+x}–ZrO{sub 2}–FeO{sub y} corium pool can form. In this case ablation of the RPV steel interacting with the molten corium is a corrosion process. Experiments carried out within the International Scientific and Technology Center's (ISTC) METCOR Project have shown that the corrosion rate can vary and depends on both surface temperature of the RPV steel and oxygen potential of the melt. If the oxygen potential is low, the corrosion rate is controlled by the solid phase diffusion of Fe ions in the corrosion layer. At high oxygen potential and steel surface layer temperature of 1050 °C and higher, the corrosion rate intensifies because of corrosion layer liquefaction and liquid phase diffusion of Fe ions. The paper analyzes conditions under which corrosion intensification occurs and can impact on in-vessel melt retention (IVR)

Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

International Nuclear Information System (INIS)

Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

2004-01-01

The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

2016-04-17

A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

In-Vessel Retention Modeling Capabilities of SCDAP/RELAP5-3DC

International Nuclear Information System (INIS)

Knudson, D.L.; Rempe, J.L.

2002-01-01

Molten core materials may relocate to the lower head of a reactor vessel in the latter stages of a severe accident. Under such circumstances, in-vessel retention (IVR) of the molten materials is a vital step in mitigating potential severe accident consequences. Whether IVR occurs depends on the interactions of a number of complex processes including heat transfer inside the accumulated molten pool, heat transfer from the molten pool to the reactor vessel (and to overlying fluids), and heat transfer from exterior vessel surfaces. SCDAP/RELAP5-3D C has been developed at the Idaho National Engineering and Environmental Laboratory to facilitate simulation of the processes affecting the potential for IVR, as well as processes involved in a wide variety of other reactor transients. In this paper, current capabilities of SCDAP/RELAP5-3D C relative to IVR modeling are described and results from typical applications are provided. In addition, anticipated developments to enhance IVR simulation with SCDAP/RELAP5-3D C are outlined. (authors)

In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

Energy Technology Data Exchange (ETDEWEB)

F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

2005-01-01

In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

In-vessel tritium retention and removal in ITER-FEAT

International Nuclear Information System (INIS)

Federici, G.; Brooks, J.N.; Iseli, M.; Wu, C.H.

2001-01-01

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ∝350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

In-vessel tritium retention and removal in ITER-FEAT

Energy Technology Data Exchange (ETDEWEB)

Federici, G. [ITER Garching Joint Work Site, Garching (Germany); Brooks, J.N. [Argonne National Lab., IL (United States); Iseli, M. [ITER Naka Joint Work Site, Naka-gun (Japan); Wu, C.H. [EFDA Close Support Unit, Garching (Germany)

2001-07-01

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruption, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., {proportional_to}350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R and D work is required to narrow the remaining uncertainties are also briefly discussed. (orig.)

In-Vessel Tritium Retention and Removal in ITER-FEAT

Science.gov (United States)

Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

Adoption of in-vessel retention concept for VVER-440/V213 reactors in Central European Countries

Energy Technology Data Exchange (ETDEWEB)

Matejovic, Peter, E-mail: peter.matejovic@ivstt.sk [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Barnak, Miroslav; Bachraty, Milan; Vranka, Lubomir [Inzinierska Vypoctova Spolocnost (IVS), Jana Holleho 5, 91701 Trnava (Slovakia); Berky, Robert [Integrita a Bezpecnost Ocelovych Konstrukcii, Rybnicna 40, 831 07 Bratislava (Slovakia)

2017-04-01

Highlights: • Design of in-vessel retention concept for VVER-440/V213 reactors. • Thermal loads acting on the inner reactor surface. • Structural response of reactor pressure vessel. • External reactor vessel cooling. - Abstract: An in-vessel retention (IVR) concept was proposed for standard VVER-440/V213 reactors equipped with confinement made of reinforced concrete and bubbler condenser pressure suppression system. This IVR concept is based on simple modifications of existing plant technology and thus it was attractive for plant operators in Central European Countries. Contrary to the solution that was adopted before at Loviisa NPP in Finland (two units of VVER-440/V213 reactor with steel confinement equipped with ice condenser), the coolant access to the reactor pressure vessel from flooded cavity is enabled via closable hole installed in the centre of thermal shield of the reactor lower head instead of lowering this massive structure in the case of severe accident. As a consequence, the crucial point of this IVR concept is narrow gap between torispherical lower head and thermal and biological shield. Here the highest thermal flux is expected in the case of severe accident. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width for coolant flow are of primarily importance. In this contribution the attention is paid especially to the analytical support with emphasis to the following points: 1) {sup ∗}Estimation of thermal loads acting on the inner reactor surface; 2) {sup ∗}Estimation of structural response of reactor pressure vessel (RPV) with emphasis on the deformation of outer reactor surface and its impact on the annular gap between RPV wall and thermal/biological shield; 3) {sup ∗}Analysis of external reactor vessel cooling. For this purpose the ASTEC code was used for performing analysis of core degradation scenarios, the ANSYS code for structural analysis of reactor vessel

Testing plan for critical heat flux measurement during in-vessel retention

International Nuclear Information System (INIS)

Aoki, Kazuyoshi; Iwaki, Chikako; Sato, Hisaki; Mimura, Satoshi; Kanamori, Daisuke

2015-01-01

In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations. (author)

Assessment of in-vessel corium retention in CPR1000

International Nuclear Information System (INIS)

Chen Xing; Zhang Shishun; Lin Jiming

2011-01-01

The In-Vessel corium Retention (IVR) strategy of Chinese 1000 MW class commercial pressurized water reactor (CPR1000) is assessed by Risk-Oriented Accident Analysis Methodology (ROAAM). Four representative severe accident scenarios are selected for the IVR assessment in this paper. According to four representative severe accident scenarios consequence calculated by the deterministic code combined with engineering judgment, the input probability distribution of the assessment is determined. Success probability of IVR from the viewpoint of thermal failure is then predicted using MOPOL code. MOPOL is a code developed basing on the well known ROAAM frame and heat transfer model of corium. It is demonstrated that the success probability of IVR by Reactor Cavity Flooding in CPR1000 is potentially higher than 99%. Application of IVR strategy in CPR1000 is envisioned probable if a further more comprehensive risk-benefit evaluation conclusion is positive. (authors)

Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

Energy Technology Data Exchange (ETDEWEB)

Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

Advanced Approach of Reactor Pressure Vessel In-service Inspection

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Pajnic, M.

2006-01-01

The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

Large-scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the ''flooded cavity'', is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Large-Scale testing of in-vessel debris cooling through external flooding of the reactor pressure vessel in the CYBL facility

International Nuclear Information System (INIS)

Chu, T.Y.; Bentz, J.H.; Bergeron, K.D.; Slezak, S.E.; Simpson, R.B.

1994-01-01

The possibility of achieving in-vessel core retention by flooding the reactor cavity, or the open-quotes flooded cavityclose quotes, is an accident management concept currently under consideration for advanced light water reactors (ALWR), as well as for existing light water reactors (LWR). The CYBL (CYlindrical BoiLing) facility is a facility specifically designed to perform large-scale confirmatory testing of the flooded cavity concept. CYBL has a tank-within-a-tank design; the inner 3.7 m diameter tank simulates the reactor vessel, and the outer tank simulates the reactor cavity. The energy deposition on the bottom head is simulated with an array of radiant heaters. The array can deliver a tailored heat flux distribution corresponding to that resulting from core melt convection. The present paper provides a detailed description of the capabilities of the facility, as well as results of recent experiments with heat flux in the range of interest to those required for in-vessel retention in typical ALWRs. The paper concludes with a discussion of other experiments for the flooded cavity applications

Hydrogen Gas Retention and Release from WTP Vessels: Summary of Preliminary Studies

Energy Technology Data Exchange (ETDEWEB)

Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bontha, Jagannadha R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mahoney, Lenna A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rassat, Scot D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bao, Jie [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Boeringa, Gregory K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buchmiller, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Burns, Carolyn A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chun, Jaehun [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Karri, Naveen K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Li, Huidong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Tran, Diana N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2015-07-01

The Hanford Waste Treatment and Immobilization Plant (WTP) is currently being designed and constructed to pretreat and vitrify a large portion of the waste in the 177 underground waste storage tanks at the Hanford Site. A number of technical issues related to the design of the pretreatment facility (PTF) of the WTP have been identified. These issues must be resolved prior to the U.S. Department of Energy (DOE) Office of River Protection (ORP) reaching a decision to proceed with engineering, procurement, and construction activities for the PTF. One of the issues is Technical Issue T1 - Hydrogen Gas Release from Vessels (hereafter referred to as T1). The focus of T1 is identifying controls for hydrogen release and completing any testing required to close the technical issue. In advance of selecting specific controls for hydrogen gas safety, a number of preliminary technical studies were initiated to support anticipated future testing and to improve the understanding of hydrogen gas generation, retention, and release within PTF vessels. These activities supported the development of a plan defining an overall strategy and approach for addressing T1 and achieving technical endpoints identified for T1. Preliminary studies also supported the development of a test plan for conducting testing and analysis to support closing T1. Both of these plans were developed in advance of selecting specific controls, and in the course of working on T1 it was decided that the testing and analysis identified in the test plan were not immediately needed. However, planning activities and preliminary studies led to significant technical progress in a number of areas. This report summarizes the progress to date from the preliminary technical studies. The technical results in this report should not be used for WTP design or safety and hazards analyses and technical results are marked with the following statement: “Preliminary Technical Results for Planning – Not to be used for WTP Design

In-vessel coolability and retention of a core melt. Volume 1

Energy Technology Data Exchange (ETDEWEB)

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T. [California Univ., Santa Barbara, CA (United States). Center for Risk Studies and Safety

1996-10-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided.

In-vessel coolability and retention of a core melt. Volume 2

International Nuclear Information System (INIS)

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

1996-10-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided

In-vessel coolability and retention of a core melt. Volume 1

International Nuclear Information System (INIS)

Theofanous, T.G.; Liu, C.; Additon, S.; Angelini, S.; Kymaelaeinen, O.; Salmassi, T.

1996-10-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is physically unreasonable. Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings. The AP600 is particularly favorable to in-vessel retention. Some ideas to enhance the assessment basis as well as performance in this respect, for applications to larger and/or higher power density reactors are also provided

Natural Convection Heat Transfer of Oxide Pool During In-Vessel Retention of Core Melts

Energy Technology Data Exchange (ETDEWEB)

Park, Hae-Kyun; Chung, Bum-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The integrity of reactor vessel may be threatened by the heat generation at the oxide pool and to the natural convection heat transfer to the reactor vessel by those two layers. Therefore, External Reactor Vessel Cooling (ERVC) is performed in order to secure the integrity of the reactor vessel. Whether the IVR(In-Vessel Retention) Strategy can be applicable to a larger reactor is the technical concern, which nourished the research interest for the natural convection heat transfer of metal and oxide pool and ERVC performance. Especially, it is hard to simulate oxide pool by experimentally due to the high level of buoyancy. Moreover, the volumetrically exothermic working fluid should be adopted to simulate the behavior of the core melts. Therefore, the volumetric heat sources that immersed in the working fluid have been adopted to simulate oxide pool by experiment. We investigated oxide pool with two different designs of the volumetric heat sources that adopted previous experiments. The investigation was performed by mass transfer experiment using analogy between heat and mass transfers. The results were compared to previous studies. We simulated the natural convection heat transfer of the oxide pool by mass transfer experiment. The isothermally cooled condition was established by limiting current technique firstly. The results were compared to previous studies under identical design of the volumetric heat sources. The average Nu's of the curvature and the top plate were close to the previous studies.

Advanced toroidal facility vaccuum vessel stress analyses

International Nuclear Information System (INIS)

Hammonds, C.J.; Mayhall, J.A.

1987-01-01

The complex geometry of the Advance Toroidal Facility (ATF) vacuum vessel required special analysis techniques in investigating the structural behavior of the design. The response of a large-scale finite element model was found for transportation and operational loading. Several computer codes and systems, including the National Magnetic Fusion Energy Computer Center Cray machines, were implemented in accomplishing these analyses. The work combined complex methods that taxed the limits of both the codes and the computer systems involved. Using MSC/NASTRAN cyclic-symmetry solutions permitted using only 1/12 of the vessel geometry to mathematically analyze the entire vessel. This allowed the greater detail and accuracy demanded by the complex geometry of the vessel. Critical buckling-pressure analyses were performed with the same model. The development, results, and problems encountered in performing these analyses are described. 5 refs., 3 figs

Assessment of In-vessel corium retention for VVER-440/V213

International Nuclear Information System (INIS)

Matejovic, P.; Barnak, M.; Bachraty, M.; Berky, R.

2011-01-01

In-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) has been recognised as a feasible and promising severe accident management strategy for VVER-440/V213 reactors. In general, the avoiding of boiling crisis on outer (cooled) RPV (reactor pressure vessel) surface is sufficient condition for preserving the RPV integrity. The crucial point of the proposed IVR concept for VVER-440/V213 is the narrow gap between elliptical lower head and thermal and biological shield. In the cold conditions the width of this gap is only about 2 cm and would be even lower in hot IVR conditions, when the reactor wall is subjected to large thermal gradients due to temperature difference between the hot inner surface (loaded by corium) and cold outer surface (which is cooled by water in flooded cavity). Sufficient gap should remain free for coolant flow for the success of the proposed IVR concept. Thus, realistic estimation of thermal load and corresponding deformations of reactor wall and their impact on gap width are of primarily importance. Two different approaches were used for the estimation of the thermal load: a conservative approach and a transient approach, both were computed with the ASTEC code. The structural analysis of RPV subjected to IVR load was performed using the finite element method (FEM) code ANSYS release 10.0. From the results obtained it follows, that even when the RPV is subjected to limiting loading conditions during severe accident, there should be sufficient gap width (∼ 1 cm) between RPV wall and thermal/biological shield for the coolant flow in natural circulation regime alongside the outer surface of the RPV wall

Carotid artery stenosis: Performance of advanced vessel analysis software in evaluating CTA

International Nuclear Information System (INIS)

Tsiflikas, Ilias; Biermann, Christina; Thomas, Christoph; Ketelsen, Dominik; Claussen, Claus D.; Heuschmid, Martin

2012-01-01

Objectives: The aim of this study was to evaluate time efficiency and diagnostic reproducibility of an advanced vessel analysis software for diagnosis of carotid artery stenosis. Material and methods: 40 patients with suspected carotid artery stenosis received head and neck DE-CTA as part of their pre-interventional workup. Acquired data were evaluated by 2 independent radiologists. Stenosis grading was performed by MPR eyeballing with freely adjustable MPRs and with a preliminary prototype of the meanwhile available client-server and advanced visualization software syngo.via CT Vascular (Siemens Healthcare, Erlangen, Germany). Stenoses were graded according to the following 5 categories: I: 0%, II: 1–50%, III: 51–69%, IV: 70–99% and V: total occlusion. Furthermore, time to diagnosis for each carotid artery was recorded. Results: Both readers achieved very good specificity values and good respectively very good sensitivity values without significant differences between both reading methods. Furthermore, there was a very good correlation between both readers for both reading methods without significant differences (kappa value: standard image interpretation k = 0.809; advanced vessel analysis software k = 0.863). Using advanced vessel analysis software resulted in a significant time saving (p < 0.0001) for both readers. Time to diagnosis could be decreased by approximately 55%. Conclusions: Advanced vessel analysis application CT Vascular of the new imaging software syngo.via (Siemens Healthcare, Forchheim, Germany) provides a high rate of reproducibility in assessment of carotid artery stenosis. Furthermore a significant time saving in comparison to standard image interpretation is achievable

Carotid artery stenosis: Performance of advanced vessel analysis software in evaluating CTA

Energy Technology Data Exchange (ETDEWEB)

Tsiflikas, Ilias, E-mail: ilias.tsiflikas@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Biermann, Christina, E-mail: christina.biermann@siemens.com [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Siemens AG, Siemens Healthcare Consulting, Allee am Röthelheimpark 3A, 91052 Erlangen (Germany); Thomas, Christoph, E-mail: christoph.thomas@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Ketelsen, Dominik, E-mail: dominik.ketelsen@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Claussen, Claus D., E-mail: claus.claussen@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany); Heuschmid, Martin, E-mail: martin.heuschmid@med.uni-tuebingen.de [University Hospital of Tuebingen, Diagnostic and Interventional Radiology, Hoppe-Seyler-Str. 3, 72076 Tuebingen (Germany)

2012-09-15

Objectives: The aim of this study was to evaluate time efficiency and diagnostic reproducibility of an advanced vessel analysis software for diagnosis of carotid artery stenosis. Material and methods: 40 patients with suspected carotid artery stenosis received head and neck DE-CTA as part of their pre-interventional workup. Acquired data were evaluated by 2 independent radiologists. Stenosis grading was performed by MPR eyeballing with freely adjustable MPRs and with a preliminary prototype of the meanwhile available client-server and advanced visualization software syngo.via CT Vascular (Siemens Healthcare, Erlangen, Germany). Stenoses were graded according to the following 5 categories: I: 0%, II: 1–50%, III: 51–69%, IV: 70–99% and V: total occlusion. Furthermore, time to diagnosis for each carotid artery was recorded. Results: Both readers achieved very good specificity values and good respectively very good sensitivity values without significant differences between both reading methods. Furthermore, there was a very good correlation between both readers for both reading methods without significant differences (kappa value: standard image interpretation k = 0.809; advanced vessel analysis software k = 0.863). Using advanced vessel analysis software resulted in a significant time saving (p < 0.0001) for both readers. Time to diagnosis could be decreased by approximately 55%. Conclusions: Advanced vessel analysis application CT Vascular of the new imaging software syngo.via (Siemens Healthcare, Forchheim, Germany) provides a high rate of reproducibility in assessment of carotid artery stenosis. Furthermore a significant time saving in comparison to standard image interpretation is achievable.

ASTEC application to in-vessel corium retention

International Nuclear Information System (INIS)

Tarabelli, D.; Ratel, G.; Pelisson, R.; Guillard, G.; Barnak, M.; Matejovic, P.

2009-01-01

This paper summarizes the work done in the SARNET European Network of Excellence on Severe Accidents (6th Framework Programme of the European Commission) on the capability of the ASTEC code to simulate in-vessel corium retention (IVR). This code, jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fuer Anlagen und Reaktorsicherheit mbH (GRS) for simulation of severe accidents, is now considered as the European reference simulation tool. First, the DIVA module of ASTEC code is briefly introduced. This module treats the core degradation and corium thermal behaviour, when relocated in the reactor lower head. Former ASTEC V1.2 version assumed a predefined stratified molten pool configuration with a metallic layer on the top of the volumetrically heated oxide pool. In order to reflect the results of the MASCA project, improved models that enable modelling of more general corium pool configurations were implemented by the CEA (France) into the DIVA module of the ASTEC V1.3 code. In parallel, the CEA was working on ASTEC modelling of the external reactor vessel cooling (ERVC). The capability of the ASTEC CESAR circuit thermal-hydraulics to simulate the ERVC was tested. The conclusions were that the CESAR module is capable of simulating this system although some numerical and physical instabilities can occur. Developments were then made on the coupling between both DIVA and CESAR modules in close collaboration with IRSN. In specific conditions, code oscillations remain and an analysis was made to reduce the numerical part of these oscillations. A comparison of CESAR results of the SULTAN experiments (CEA) showed an agreement on the pressure differences. The ASTEC V1.2 code version was applied to IVR simulation for VVER-440/V213 reactors assuming defined corium mass, composition and decay heat. The external cooling of reactor wall was simulated by applying imposed coolant temperature and heat transfer

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected. (author)

A wall-crawling robot for reactor vessel inspection in advanced reactors

International Nuclear Information System (INIS)

Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

1994-01-01

A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected

33 CFR 401.79 - Advance notice of arrival, vessels requiring inspection.

Science.gov (United States)

2010-07-01

..., vessels requiring inspection. 401.79 Section 401.79 Navigation and Navigable Waters SAINT LAWRENCE SEAWAY... Reports § 401.79 Advance notice of arrival, vessels requiring inspection. Every vessel shall provide at... reinspection of the ship is required. [70 FR 12973, Mar. 17, 2005] ...

In-vessel core debris retention experiments. Final report

International Nuclear Information System (INIS)

1998-10-01

The in-vessel cooling experimental program (Phase 1 and 2) was motivated by the survivability of the TMI lower vessel head during the TMI-2 accident. During that accident, molten debris relocation into the water filled lower head resulted in a localized hot spot in the lower head, but no lower head failure occurred. A postulated set of mechanisms which could be involved in and responsible for the survivability of the TMI lower head were identified and experimentally investigated as part of this program. These mechanisms included: the formation of a gap (contact resistance) between the relocated and frozen debris and the vessel wall was a key aspect of the in-vessel cooling mechanism; wall heatup due to the relocated debris in the presence of wall stress due to a pressure gradient across the vessel wall; gap growth due to a lack of debris adherence to the vessel wall and material creep of the heated vessel wall; and the potential for enhanced wall cooling due to gap growth. Each of these postulated mechanisms was investigated in this experimental program. This report summarizes the several insights and conclusions that were obtained from this experimental program. This report documents the entire set of five experiments completed in Phase 2 of this experimental program. Results from the Phase 1 effort were used to plan and select the Phase 2 test matrix. Conclusions from the Phase 1 and 2 experiments are identified and recommendations for future work are provided

Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W

International Nuclear Information System (INIS)

Skinner, C.H.; Haasz, A.A.; Alimov, V.Kh.; Bekris, N.; Causey, R.A.; Clark, R.E.H.; Coad, J.P.; Davis, J.W.; Doerner, R.P.; Mayer, M.; Pisarev, A.; Roth, J.; Tanabe, T.

2008-01-01

Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described

Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W.

Energy Technology Data Exchange (ETDEWEB)

Skinner, C H; Alimov, Kh; Bekris, N; Causey, R A; Clark, R.E.H.; Coad, J P; Davis, J W; Doerner, R P; Mayer, M; Pisarev, A; Roth, J

2008-03-29

Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.

Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

Energy Technology Data Exchange (ETDEWEB)

Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

The Employment Retention and Advancement Project: Paths to Advancement for Single Parents

Science.gov (United States)

Miller, Cynthia; Deitch, Victoria; Hill, Aaron

2010-01-01

Between 2000 and 2003, the Employment Retention and Advancement (ERA) project identified and implemented a diverse set of innovative models designed to promote employment stability and wage or earnings progression among low-income individuals, mostly current or former welfare recipients. The project's goal was to determine which strategies could…

The impact of policy change on job retention and advancement

OpenAIRE

Richard Dickins; Abigail McKnight

2008-01-01

This paper examines the impact of the Working Families Tax Credit (WFTC) on employment retention and advancement. The WFTC, which replaced Family Credit in October 1999, supplemented earnings of low paid workers living in low income families. It was designed to increase the financial incentive for low skilled workers to find and remain in work and in the process boost their family income. It finds evidence that WFTC increased employment retention among male recipients. WFTC does not appear to...

Design description of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Nelson, B.E.; Vinyard, L.M.; Williamson, D.F.

1983-01-01

The Advanced Toroidal Facility (ATF) will be a stellarator experiment to investigate improvements in toroidal confinement. The vacuum vessel for this facility will provide the appropriate evacuated region for plasma containment within the helical field (HF) coils. The vessel is designed to provide the maximum reasonable volume inside the HF coils and to provide the maximum reasonable access for future diagnostics. The vacuum vessel design is at an early phase and all of the details have not been completed. The heat transfer analysis and stress analysis completed during the conceptual design indicate that the vessel will not change drastically

Americium behaviour in plastic vessels

Energy Technology Data Exchange (ETDEWEB)

Legarda, F.; Herranz, M. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Idoeta, R., E-mail: raquel.idoeta@ehu.e [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Abelairas, A. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain)

2010-07-15

The adsorption of {sup 241}Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of {sup 241}Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of {sup 241}Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

Americium behaviour in plastic vessels

International Nuclear Information System (INIS)

Legarda, F.; Herranz, M.; Idoeta, R.; Abelairas, A.

2010-01-01

The adsorption of 241 Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of 241 Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of 241 Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

Cold source vessel development for the advanced neutron source

Energy Technology Data Exchange (ETDEWEB)

Williams, P.T.; Lucas, A.T. [Oak Ridge National Lab., TN (United States)

1995-09-01

The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory (ORNL), will be a user-oriented neutron research facility that will produce the most intense flux of neutrons in the world. Among its many scientific applications, the productions of cold neutrons is a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410 mm diameter sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design are being performed with multi-dimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This paper presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that will be used to verify the final design.

Americium behaviour in plastic vessels.

Science.gov (United States)

Legarda, F; Herranz, M; Idoeta, R; Abelairas, A

2010-01-01

The adsorption of (241)Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of (241)Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of (241)Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification. Copyright 2009 Elsevier Ltd. All rights reserved.

Overview of in-vessel retention concept involving level of passivity: with application to evolutionary pressurized water reactor design

International Nuclear Information System (INIS)

Ghyym, Seong H.

1998-01-01

In this work, one strategy of severe accident management, the applicability of the in-vessel retention (IVR) concept, which has been incorporated in passive type reactor designs, to evolutionary type reactor designs, is examined with emphasis on the method of external reactor vessel cooling (ERVC) to realize the IVR concept in view of two aspects: for the regulatory aspect, it is addressed in the context of the resolution of the issue of corium coolability; for the technical one, the reliance on and the effectiveness of the IVR concept are mentioned. Additionally, for the ERVC method to be better applied to designs of the evolutionary type reactor, the conditions to be met are pointed out in view of the technical aspect. Concerning the issue of corium coolability/quenchability, based on results of the review, plausible alternative strategies are proposed. According to the decision maker's risk behavior, these would help materialize the conceptual design for evolutionary type reactors, especially Korea Next Generation Reactors (KNGRs), which have been developing at the Korea Electric Power Research Institute (KEPRI): (A1) Strategy 1A: strategy based on the global approach using the reliance on the wet cavity method; (A2) Strategy 1B: strategy based on the combined approach using both the reliance on the wet cavity method and the counter-measures for preserving containment integrity; (A3) Strategy 2A: strategy based on the global approach to the reliance on the ERVC method; (A4) Strategy 2B: strategy based on the balanced approach using both the reliance on the ERVC method and the countermeasures for preserving containment integrity. Finally, in application to an advanced pressurized water reactor (PWR) design, several recommendations are made in focusing on both monitoring the status of approaches and preparing countermeasures in regard to the regulatory and the technical aspects

Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Kim, Sung Joong; Seo, Seungwon; Lee, Seongnyeon; KIm, Hwan Yeol; Ha, Kwang Soon; Park, Jonghwa; Park, Raejoon

2013-01-01

The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters

Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Joong; Seo, Seungwon; Lee, Seongnyeon [Hanyang Univ., Seoul (Korea, Republic of); KIm, Hwan Yeol; Ha, Kwang Soon; Park, Jonghwa; Park, Raejoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters.

Advanced nickel/hydrogen dependent pressure vessel (DPV) cell and battery concepts

Energy Technology Data Exchange (ETDEWEB)

Caldwell, D.B. [Technologies Div., Eagle Picher Industries, Inc., Joplin, MO (United States); Fox, C.L. [Technologies Div., Eagle Picher Industries, Inc., Joplin, MO (United States); Miller, L.E. [Technologies Div., Eagle Picher Industries, Inc., Joplin, MO (United States)

1997-03-01

The dependent pressure vessel (DPV) nickel/hydrogen (NiH{sub 2}) design is being developed by Eagle-Picher industries, Inc. (EPI) as an advanced battery for military and commercial aerospace and terrestrial applications. The DPV cell design offers high specific energy and energy density as well as reduced cost, while retaining the established individual pressure vessel (IPV) technology, flight heritage and database. This advanced DPV design also offers a more efficient mechanical, electrical and thermal cell and battery configuration and a reduced parts count. The DPV battery design promotes compact, minimum volume packaging and weight efficiency, and delivers cost and weight savings with minimal design risks. (orig.)

FE-simulation of the viscoplastic behaviour of different RPV steels in the frame of in-vessel melt retentions scenarios

International Nuclear Information System (INIS)

Altstadt, E.; Willschuetz, H.G.; Mueller, G.

2004-01-01

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of REactor VEssel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. The geometrical scale of the experiments is 1:10 compared to a common LWR. This paper deals with the experimental, numerical, and metallographical results of the creep failure experiment EC-FOREVER-4, where the American pressure vessel steel SA533B was applied for the lower head. For comparison the results of the experiment EC-FOREVER-3B, build of the French 16MND5 steel, are discussed, too. Emphasis is put on the differences in the viscoplastic behaviour of different heats of the RPV steel. For this purpose, the creep tests in the frame of the LHF/OLHF experiments are reviewed, too. As a hypothesis it is stated that the sulphur content could be responsible for differences in the creep behaviour. (orig.)

Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

International Nuclear Information System (INIS)

Chipley, K.K.; Frey, G.N.

1985-01-01

The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

Smooth muscle cell recruitment to lymphatic vessels requires PDGFB and impacts vessel size but not identity.

Science.gov (United States)

Wang, Yixin; Jin, Yi; Mäe, Maarja Andaloussi; Zhang, Yang; Ortsäter, Henrik; Betsholtz, Christer; Mäkinen, Taija; Jakobsson, Lars

2017-10-01

Tissue fluid drains through blind-ended lymphatic capillaries, via smooth muscle cell (SMC)-covered collecting vessels into venous circulation. Both defective SMC recruitment to collecting vessels and ectopic recruitment to lymphatic capillaries are thought to contribute to vessel failure, leading to lymphedema. However, mechanisms controlling lymphatic SMC recruitment and its role in vessel maturation are unknown. Here, we demonstrate that platelet-derived growth factor B (PDGFB) regulates lymphatic SMC recruitment in multiple vascular beds. PDGFB is selectively expressed by lymphatic endothelial cells (LECs) of collecting vessels. LEC-specific deletion of Pdgfb prevented SMC recruitment causing dilation and failure of pulsatile contraction of collecting vessels. However, vessel remodelling and identity were unaffected. Unexpectedly, Pdgfb overexpression in LECs did not induce SMC recruitment to capillaries. This was explained by the demonstrated requirement of PDGFB extracellular matrix (ECM) retention for lymphatic SMC recruitment, and the low presence of PDGFB-binding ECM components around lymphatic capillaries. These results demonstrate the requirement of LEC-autonomous PDGFB expression and retention for SMC recruitment to lymphatic vessels, and suggest an ECM-controlled checkpoint that prevents SMC investment of capillaries, which is a common feature in lymphedematous skin. © 2017. Published by The Company of Biologists Ltd.

Heat load imposed on reactor vessels during in-vessel retention of core melts

Energy Technology Data Exchange (ETDEWEB)

Kim, Su-Hyeon; Chung, Bum-Jin, E-mail: bjchung@khu.ac.kr

2016-11-15

Highlights: • Angular heat load on reactor vessel by natural convection of oxide pool was measured. • High Ra was achieved by using mass transfer experiments based on analogy concept. • Measured Nusselt numbers agreed reasonably with the other existing studies. • Three different types of volumetric heat sources were compared. • They didn’t affect the heat flux of the top plate but affected those of the reactor vessel. - Abstract: We measured the heat load imposed on reactor vessels by natural convection of the oxide pool in severe accidents. Based on the analogy between heat and mass transfer, mass transfer experiments were performed using a copper sulfate electroplating system. A modified Rayleigh number of the order 10{sup 14} was achieved in a small facility with a height of 0.1 m. Three different types of volumetric heat sources were compared and the average Nusselt number of the curved surface was 39% lower, whereas in the case of the top plate was 6% higher than in previous studies with a two-dimensional geometry due to the high Sc value of this study. Reliable experimental data on the angular heat flux ratios were reported compared to those of the BALI and SIGMA CP facilities in terms of fluctuations and consistency.

Proceedings of the 1985 pressure vessels and piping conference. Volume PVP-98-8. Fracture, fatigue and advanced mechanics

International Nuclear Information System (INIS)

Short, W.E.; Zamrik, S.Y.

1985-01-01

State-of-the-art engineering practices in pressure vessel and piping technology are the result of continual efforts in the evaluation of problems which have been experienced and the development of appropriate design and analysis methods for those applications. The resulting advances in technology benefit industry with properly engineered, safe, cost-effective pressure vessels and piping systems. To this end, advanced study continues in specialized areas of mechanical engineering such as fracture mechanics, experimental stress analysis, high pressure applications and related material considerations, as well as advanced techniques for evaluation of commonly encountered design problems. This volume is comprised of current technical papers on various aspects of fracture, fatigue and advanced mechanics as related to the design and analysis of pressure vessels and piping

Effect of in-core instrumentation mounting location on external reactor vessel cooling

International Nuclear Information System (INIS)

Suh, Jungsoo; Ha, Huiun

2017-01-01

Highlights: • Numerical simulations were conducted for the evaluation of an IVR-ERVC application. • The ULPU-V experiment was simulated for the validation of numerical method. • The effect of ICI mounting location on an IVR-ERVC application was investigated. • TM-ICI is founded to be superior to BM-ICI for successful application of IVR-ERVC. - Abstract: The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

Effectiveness of External Reactor Vessel Cooling (ERVC) strategy for APR1400 and issues of phenomenological uncertainties

International Nuclear Information System (INIS)

Oh, S.J.; Kim, H.T.

2007-01-01

The APR1400(Advanced Power Reactor 1400) is an evolutionary advanced light water reactor with rated thermal power of 4000 MWt. For APR1400, External Reactor Vessel Cooling (ERVC) is adopted as a primary severe accident management strategy for in-vessel retention (IVR) of corium. The ERVC is a method of IVR by submerging the reactor vessel exterior. At the early stage of the APR1400 design, only ex-vessel cooling, cooling of the core melt outside the vessel after vessel is breached, is considered based on the EPRI Utility Requirement Document for Evolutionary LWR. However, based on the progress in implementation of Severe Accident Management Guidance (SAMG) for operating plants, as well as the research findings related to ERVC, ERVC strategy is adopted as a part of key severe accident management strategies. To improve its success, the strategy is reviewed and we implemented necessary design arrangement to increase its usefulness in managing the severe accident. In this paper, we examine the evolution of ERVC concept and its implementation in APR1400. Then, we review possible approach, including Risk-Oriented Accident Analysis Methodology (ROAAM), to evaluate the effectiveness of the strategy. (authors)

In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

Energy Technology Data Exchange (ETDEWEB)

Heel, A.M.J.M. van

1995-07-01

An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP).

In-vessel core debris retention through external flooding of the reactor pressure vessel. State-of-the-art report

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-07-01

An overview of the state-of-the-art knowledge on the ex-vessel flooding accident management strategy for severe accidents in a NPP has been given. The feasibility has been discussed, as well as the in- and ex-vessel phenomena, which influence the structural integrity of the vessel. Finally, some computer codes with the ability to model the phenomena involved in ex-vessel flooding have been discussed. (orig./HP)

Design of Hemispherical Downward-Facing Vessel for Critical Heat Flux Experiment

International Nuclear Information System (INIS)

Hwang, J. S.; Suh, K. Y.

2009-01-01

The in-vessel retention (IVR) is one of major severe accident management strategies adopted by some operating nuclear power plants during a severe accident. The recent Shin-Gori Units 3 and 4 of the Advanced Power Reactor 1400 MWe (APR1400) have adopted the external reactor vessel cooling (ERVC) by reactor cavity flooding as major severe accident management strategy. The ERVC in the APR1400 design resorts to active flooding system using thermal insulator. The Corium Attack Stopper Apparatus Spherical Channel (CASA SC) tests are conducted to measure the critical power and critical heat flux (CHF) on a downward hemispherical vessel scaled down from the APR1400 lower head by 1/10 on a linear scale. CASA is designed through scaling and thermal analysis to simulate the APR1400 vessel and thermal insulator. The heated vessel of CASA SC represents the external surface of a hemisphere submerged vessel in water. The heated vessel plays an important role in the ERVC experiment depending on the configuration of oxide pool and metallic layer. Hand calculation and computational analysis are performed to produce high heat flux from the downward facing hemisphere in excess of 1 MW/m 2

An experimental study on in-vessel debris retention through gap cooling

International Nuclear Information System (INIS)

Kang, Kyung Ho; Kim, Jong Whan; Cho, Young Ro; Chang, Young Cho; Park, Rae Jun; Gu, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

1999-04-01

LAVA experiments have been performed using high temperature molten material to be relocated into the 1/8 linear scaled vessel of a reactor lower plenum filled with water. An Al 2 O 3 /Fe tehrmite melt (Al 2 O 3 only) was used as a corium simulant. In this study, the influence of various initial conditions, such as internal pressure load across the vessel wall, the material composition of the melt simulant, water subcooling and depth on gap formation were investigated. As well, the thermal and mechanical behaviors of the vessel were examined. In case the internal pressure load was imposed, the gap formation between the continuous solidified debris and the vessel wall was clearly shown with post-test examination. On the other hand, in case the internal pressure load was not imposed, the iron welded to the inner surface of the vessel and the vessel experienced ablation to about 5 mm. The cooling rate of the vessel was very slow in the tests using Al 2 O 3 /Fe thermite melt but it was rather fast using Al 2 O 3 melt. It is postulated that in the Al 2 O 3 /Fe thermite tests, the iron melt layer is so dense that water ingression into the gap is difficult due to the high pressure of escaping steam. On the other hand, in the porosity of an Al 2 O 3 melt layer could enhance water ingression into the gap by giving the flow path of the evaporated steam through the porous media. The water height and subcooling could affect the melt pool formation and the initial thermal attack to the vessel. However, at present stage, the effect of water subcooling on the thermal behavior of the vessel couldn't be generalized. For clear confirmation of the effect of water subcooling, the tests will be performed at the saturated water condition. (author). 19 refs., 3 tabs., 36 figs

The Impact of a Junior Faculty Fellowship Award on Academic Advancement and Retention.

Science.gov (United States)

Connelly, Maureen T; Sullivan, Amy M; Chinchilla, Manuel; Dale, Margaret L; Emans, S Jean; Nadelson, Carol Cooperman; Notman, Malkah Tolpin; Tarbell, Nancy J; Zigler, Corwin M; Shore, Eleanor G

2017-08-01

Academic faculty experience barriers to career development and promotion. In 1996, Harvard Medical School (HMS) initiated an intramural junior faculty fellowship to address these obstacles. The authors sought to understand whether receiving a fellowship was associated with more rapid academic promotion and retention. Junior faculty fellowship recipients and all other instructor and assistant professors at HMS between 1996 and 2011 were identified. Using propensity score modeling, the authors created a matched comparison group for the fellowship recipients based on educational background, training, academic rank, department, hospital affiliation, and demographics. Time to promotion and time to leaving were assessed by Kaplan-Meier curves. A total of 622 junior faculty received fellowships. Faculty who received fellowships while instructors (n = 480) had shorter times to promotion to assistant professor (P Women instructors advanced more quickly than matched controls, while male instructors' rates of promotions did not differ. Fellowships to support junior faculty were associated with shorter times to promotion for instructors and more sustained faculty retention for both instructors and assistant professors. This suggests that relatively small amounts of funding early in faculty careers can play a critical role in supporting academic advancement and retention.

Structural analysis and manufacture for the vacuum vessel of experimental advanced superconducting tokamak (EAST) device

International Nuclear Information System (INIS)

Song Yuntao; Yao Damao; Wu Songata; Weng Peide

2006-01-01

The experimental advanced superconducting tokamak (EAST) is an advanced steady-state plasma physics experimental device, which has been approved by the Chinese government and is being constructed as the Chinese national nuclear fusion research project. The vacuum vessel, that is one of the key components, will have to withstand not only the electromagnetic force due to the plasma disruption and the Halo current, but also the pressure of boride water and the thermal stress due to the 250 deg. C baking out by the hot pressure nitrogen gas, or the 100 deg. C hot wall during plasma operation. This paper is a report of the mechanical analyses of the vacuum vessel. According to the allowable stress criteria of American Society of Mechanical Engineers, Boiler and Pressure Vessel Committee (ASME), the maximum integrated stress intensity on the vacuum vessel is 396 MPa, less than the allowable design stress intensity 3S m (441 MPa). At the same time, some key R and D issues are presented, which include supporting system, bellows and the assembly of the whole vacuum vessel

Development of advanced manufacturing technologies for low cost hydrogen storage vessels

Energy Technology Data Exchange (ETDEWEB)

Leavitt, Mark [Quantum Fuel Systems Technologies Worldwide, Inc., Irvine, CA (United States); Lam, Patrick [Boeing Research and Technology (BR& T), Seattle, WA (United States)

2014-12-29

The U.S. Department of Energy (DOE) defined a need for low-cost gaseous hydrogen storage vessels at 700 bar to support cost goals aimed at 500,000 units per year. Existing filament winding processes produce a pressure vessel that is structurally inefficient, requiring more carbon fiber for manufacturing reasons, than would otherwise be necessary. Carbon fiber is the greatest cost driver in building a hydrogen pressure vessel. The objective of this project is to develop new methods for manufacturing Type IV pressure vessels for hydrogen storage with the purpose of lowering the overall product cost through an innovative hybrid process of optimizing composite usage by combining traditional filament winding (FW) and advanced fiber placement (AFP) techniques. A numbers of vessels were manufactured in this project. The latest vessel design passed all the critical tests on the hybrid design per European Commission (EC) 79-2009 standard except the extreme temperature cycle test. The tests passed include burst test, cycle test, accelerated stress rupture test and drop test. It was discovered the location where AFP and FW overlap for load transfer could be weakened during hydraulic cycling at 85°C. To design a vessel that passed these tests, the in-house modeling software was updated to add capability to start and stop fiber layers to simulate the AFP process. The original in-house software was developed for filament winding only. Alternative fiber was also investigated in this project, but the added mass impacted the vessel cost negatively due to the lower performance from the alternative fiber. Overall the project was a success to show the hybrid design is a viable solution to reduce fiber usage, thus driving down the cost of fuel storage vessels. Based on DOE’s baseline vessel size of 147.3L and 91kg, the 129L vessel (scaled to DOE baseline) in this project shows a 32% composite savings and 20% cost savings when comparing Vessel 15 hybrid design and the Quantum

In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

International Nuclear Information System (INIS)

Cognet, C.; Gandrille, P.

1999-01-01

In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

Basic Boiling Experiments with An Inclined Narrow Gap Associated With In-Vessel Retention

International Nuclear Information System (INIS)

Terazu, Kuninobu; Watanabe, Fukashi; Iwaki, Chikako; Yokobori, Seiichi; Akinaga, Makoto; Hamazaki, Ryoichi; SATO, Ken-ichi

2002-01-01

In the case of a severe accident with relocation of the molten corium into the lower plenum of reactor pressure vessel (RPV), the successful in-vessel corium retention (IVR) can prevent the progress to ex-vessel events with uncertainties and avoid the containment failure. One of the key phenomena governing the possibility of IVR would be the gap formation and cooling between a corium crust and the RPV wall, and for the achievement of IVR, it would be necessary to supply cooling water to RPV as early as possible. The BWR features relative to IVR behavior are a deep and massive water pool in the lower plenum, and many of control rod drive guide tubes (CRDGT) installed in the lower head of RPV, in which water is injected continuously except in the case of station blackout scenario. The present paper describes the basic boiling experiment conducted in order to investigate the boiling characteristics in an inclined narrow gap simulating a part of the lower head curvature. The boiling experiments were composed of visualization tests and heat transfer tests. In the visualization tests, two types of inclined gap were constructed using the parallel plate and the V-shaped parallel plate with heating from the top plate, and the boiling flow pattern was observed with various gap width and heat flux. These observation results showed that water was easily supplied from the gap bottom of parallel plate even in a very narrow gap with smaller width than 1 mm, and water could flow continuously in the narrow gap by the geometric and thermal imbalance from the experiment results using the V-shaped parallel plate. In the heat transfer tests, the critical heat flux (CHF) data in an inclined narrow channel formed by the parallel plates were measured in terms of the parameters of gap width, heated length and inclined angle of a channel, and the effect of inclination was incorporated into the existing CHF correlation for a narrow gap. The CHF correlation modified for an inclined narrow gap

Effect of In-Vessel Retention Strategies under Postulated SGTR Accidents of OPR1000

Energy Technology Data Exchange (ETDEWEB)

Choi, Wonjun; Lee, Yongjae; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Kim, Hwan-Yeol; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this study, MELCOR code was used to simulate the severe accident of the OPR1000. MELCOR code is computer code which enables to simulate the progression of the severe accident for light water reactors. It has been developed by Sandia National Laboratories for plant risk assessment and source term analysis since 1982. According to the probabilistic safety analysis (PSA) Level 1 of OPR1000, typical severe accident scenarios of high probability of a transition to severe accident for OPR1000 were identified as Small Break Loss of Coolant Accident (SBLOCA), Station Black out (SBO), Total Loss of Feed Water (TLOFW), and Steam Generator Tube Rupture. While the first three accidents are expected to result in the generation and transportation of the radioactive nuclides within the containment building as consequence of the core damage and subsequent reactor pressure vessel (RPV) failure, the latter accident scenario may be progressed with possible direct release of the radioactive nuclides to the environment by bypassing the containment building. Thus it is of significance to investigate the SGTR accident with a sophisticated severe accident code. This code can simulate the whole phenomena of a severe accident such as thermal-hydraulic response, core heat-up, oxidation and relocation, and fission product release and transport. Thus many researchers have used MELCOR in severe accident studies. In this study, in-vessel retention strategies were applied for postulated SGTR accidents. Mitigation effect and adverse effect of in-vessel strategies was studied in aspect of RPV failure, fission product release and containment thermal-hydraulic and hydrogen behavior. Base case of SGTR accident and three mitigation cases were simulated using MELCOR code 1.8.6. For each mitigation cases, mitigation effect and adverse effect were investigated. Conclusions can be summarized as follows: (1) RPV failure of SGTR base case occurred at 5.62 hours and fission product of RCS released to

Advanced dependent pressure vessel (DPV) nickel-hydrogen spacecraft battery design

Energy Technology Data Exchange (ETDEWEB)

Coates, D.K.; Grindstaff, B.; Swaim, O.; Fox, C. [Eagle-Picher Industries, Inc., Joplin, MO (United States). Advanced Systems Operation

1995-12-31

The dependent pressure vessel (DPV) nickel-hydrogen (NiH{sub 2}) battery is being developed as a potential spacecraft battery design for both military and commercial satellites. The limitations of standard NiH{sub 2} individual pressure vessel (IPV) flight battery technology are primarily related to the internal cell design and the battery packaging issues associated with grouping multiple cylindrical cells. The DPV cell design offers higher energy density and reduced cost, while retaining the established IPV technology flight heritage and database. The advanced cell design offers a more efficient mechanical, electrical and thermal cell configuration and a reduced parts count. The geometry of the DPV cell promotes compact, minimum volume packaging and weight efficiency. The DPV battery design offers significant cost and weight savings advantages while providing minimal design risks.

Reactor vessel supported by flexure member

International Nuclear Information System (INIS)

Crawford, J.D.; Pankow, B.

1977-01-01

According to the present invention there is provided an improved arrangement for supporting a reactor vessel within a containment structure against static and dynamic vertical loadings capable of being imposed as a result of a serious accident as well as during periods of normal plant operation. The support arrangement of the invention is, at the same time, capable of accommodating radial displacements that normally occur between the reactor vessel and the containment structure due to operational transients. The arrangement comprises a plurality of vertical columns connected between the reactor vessel and a support base within the containment structure. The columns are designed to accommodate relative displacements between the vessel and the containment structure by flexing. This eliminates the need for relative sliding movements and thus enables the columns to be securely fixed to the vessel. This elimination of a provision for relative sliding movements avoids the spaces or gaps between the retention members and the retained elements as occurred in prior art arrangements and, concomitantly, the danger of establishing impact forces on the retention members in the event of an accident is reduced. (author)

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

Energy Technology Data Exchange (ETDEWEB)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

2001-03-15

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

International Nuclear Information System (INIS)

Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin; Park, Jae Hong

2001-03-01

In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report

Applications of nano-fluids to enhance LWR accidents management in in-vessel retention and emergency core cooling systems

International Nuclear Information System (INIS)

Chupin, A.; Hu, L. W.; Buongiorno, J.

2008-01-01

Water-based nano-fluid, colloidal dispersions of nano-particles in water; have been shown experimentally to increase the critical heat flux and surface wettability at very low concentrations. The use of nano-fluids to enhance accidents management would allow either to increase the safe margins in case of severe accidents or to upgrade the power of an existing power plant with constant margins. Building on the initial work, computational fluid dynamics simulations of the nano-fluid injection system have been performed to evaluate the feasibility of a nano-fluid injection system for in-vessel retention application. A preliminary assessment was also conducted on the emergency core cooling system of the European Pressurized Reactor (EPR) to implement a nano-fluid injection system for improving the management of loss of coolant accidents. Several design options were compared/or their respective merits and disadvantages based on criteria including time to injection, safety impact, and materials compatibility. (authors)

MS PHD'S: Effective Strategies for the Retention and Advancement of URM Students in ESS

Science.gov (United States)

Escalera, J.; Burgess, A. K.; Pace, L.; Scott, O.; Strickland, J.; Johnson, A.; Williamson Whitney, V.; Ithier-Guzman, W.

2012-12-01

The Minorities Striving and Pursuing Higher Degrees of Success (MS PHD'S) Professional Development Program in Earth system science (ESS) is a model initiative for improving the retention of underrepresented minority (URM) students in STEM fields. Entering its ninth cohort, MS PHD'S remains committed to helping URM undergraduate and graduate students achieve outstanding careers in ESS. MS PHD'S facilitates URM student achievement through a three-phase program designed to increase student exposure to the ESS community. By engaging in a series of professional development and skill building exercises, peer-to-peer community building activities, participation in scientific society conferences and workshops, mentoring by URM and other scientists, and a virtual community, URM students gain the confidence and support necessary to achieve their academic goals and enter the ESS workforce. Since its inception, MS PHD'S continues to support 189 participants. Of these 189 participants, 35 have advanced from undergraduate and graduate academic pathways to completion of their PhD and another 60 are currently enrolled in doctoral programs. MS PHD'S maintains close ties with program alumni to further support retention, inclusivity, and broadening participation of URM students and graduates in STEM activities. Its model is built on reengaging alumni to become mentors and leaders for each new cohort as well as facilitating valuable opportunities for alumni to advance in their ESS related academic and professional career pathways.

Prestressed-concrete pressure vessels and their applicability to advanced-energy-system concepts

International Nuclear Information System (INIS)

Naus, D.J.

1983-01-01

Prestressed concrete pressure vessels (PCPVs) are, in essence, spaced steel structures since their strength is derived from a multitude of steel elements made up of deformed reinforcing bars and prestressing tendons which are present in sufficient quantities to carry tension loads imposed on the vessel. Other major components of a PCPV include the concrete, liner and cooling system, and insulation. PCPVs exhibit a number of advantages which make them ideally suited for application to advanced energy concepts: fabricability in virtually any size and shape using available technology, improved safety, reduced capital costs, and a history of proven performance. PCPVs have many applications to both nuclear- and non-nuclear-based energy systems concepts. Several of these concepts will be discussed as well as the research and development activities conducted at ORNL in support of PCPV development

Nuclear reactor support and seismic restraint with in-vessel core retention cooling features

Energy Technology Data Exchange (ETDEWEB)

Edwards, Tyler A.; Edwards, Michael J.

2018-01-23

A nuclear reactor including a lateral seismic restraint with a vertically oriented pin attached to the lower vessel head and a mating pin socket attached to the floor. Thermally insulating materials are disposed alongside the exterior surface of a lower portion of the reactor pressure vessel including at least the lower vessel head.

Changes in cross-sectional area of pulmonary vessels on chest computed tomography after chemotherapy in patients with advanced non-squamous non-small-cell lung cancer.

Science.gov (United States)

Karayama, Masato; Inui, Naoki; Kusagaya, Hideki; Suzuki, Seiichiro; Inoue, Yusuke; Enomoto, Noriyuki; Fujisawa, Tomoyuki; Nakamura, Yutaro; Suda, Takafumi

2016-05-01

Chemotherapy is associated with a risk of vascular damage. Novel anti-angiogenic agents, which can directly affect tumor angiogenesis, are increasingly being used. However, the effects of these agents on normal vasculature are not well understood. Here, we evaluated the effects of chemotherapy in general, and the anti-angiogenic agent bevacizumab, more specifically, on the pulmonary vasculature in patients with advanced non-squamous non-small-cell lung cancer (NSCLC). For this, we used the cross-sectional area of pulmonary vessels (CSA), which is an easily measurable indicator of small pulmonary vasculature on non-contrast chest computed tomography (CT). We retrospectively reviewed CT scans of the lungs of 75 chemo-naïve patients with advanced non-squamous NSCLC, for measurement of CSA, before and after first-line platinum-based chemotherapy, using a semi-automatic image-processing program. Measured vessels were classified in two groups: small vessels with CSA area (%CSAsmall-diameter vessels, with a significant decrease in %CSAsmall pulmonary vascular damage. Use of bevacizumab does not enhance the reduction in area of pulmonary vessels.

Validating Student Satisfaction Related to Persistence, Academic Performance, Retention and Career Advancement within ODL Perspectives

Science.gov (United States)

Sembiring, Maximus Gorky

2015-01-01

Student satisfaction associated with persistence, academic performance, retention, and its relations to career advancement were examined. It was aimed at measuring service quality (Servqual) dimensions as a foundation of satisfaction and how, in what comportments, they were interrelated. The study was conducted under explanatory-design. Data was…

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

Tutu, N.K.; Ginsberg, T.; Klages, J.R.

1991-01-01

This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

Validating Student Satisfaction Related to Persistence, Academic Performance, Retention and Career Advancement within ODL Perspectives

Directory of Open Access Journals (Sweden)

Maximus Gorky Sembiring

2015-11-01

Full Text Available Student satisfaction associated with persistence, academic performance, retention, and its relations to career advancement were examined. It was aimed at measuring service quality (Servqual dimensions as a foundation of satisfaction and how, in what comportments, they were interrelated. The study was conducted under explanatory-design. Data was collected proportionally and purposively followed by congregating them through unified interviews. Population was 1,814 Universitas Terbuka students domiciled overseas; 350 questionnaires were dispersed, 169 completed. Satisfaction was assessed by examining Servqual dimensions. Importance-performance analysis (IPA and customer-satisfaction index (CSI were applied to measure satisfaction and the level of its importance. Structural equation model (SEM was then employed to examine influencing variables. Nine hypotheses developed were all validated by the analysis. Responsiveness, assurance, tangible, reliability, and empathy were in harmony to satisfaction. Career advancement, retention, academic performance, and persistence were influenced by satisfaction. Qualitative inquiry implemented afterwards was basically coherent with the quantitative findings.

Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

International Nuclear Information System (INIS)

Chu, T.Y.; Slezak, S.E.; Bentz, J.H.; Pasedag, W.F.

1994-01-01

This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm 2 across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests

Fluid Dynamics of Magnetic Nanoparticles in Simulated Blood Vessels

Science.gov (United States)

Blue, Lauren; Sewell, Mary Kathryn; Brazel, Christopher S.

2008-11-01

Magnetic nanoparticles (MNPs) can be used to locally target therapies and offer the benefit of using an AC magnetic field to combine hyperthermia treatment with the triggered release of therapeutic agents. Here, we investigate localization of MNPs in a simulated environment to understand the relationship between magnetic field intensity and bulk fluid dynamics to determine MNP retention in a simulated blood vessel. As MNPs travel through blood vessels, they can be slowed or trapped in a specific area by applying a magnetic field. Magnetic cobalt ferrite nanoparticles were synthesized and labeled with a fluorescent rhodamine tag to visualize patterns in a flow cell, as monitored by a fluorescence microscope. Particle retention was determined as a function of flow rate, concentration, and magnetic field strength. Understanding the relationship between magnetic field intensity, flow behavior and nanoparticle characteristics will aid in the development of therapeutic systems specifically targeted to diseased tissue.

State of the Art Report for the In-Vessel Late Core Melt Progression

International Nuclear Information System (INIS)

Kim, Hee Dong; Kang, Kyoung Ho; Park, Rae Joon

2009-04-01

The formation of corium pool in the reactor vessel lower head and its behavior is still an important issue. This issue is closely related to understanding of the core melting, its course, critical phases and timing during severe accidents and the influence of these processes on the accident progression, especially the evaluation of in-vessel retention by external reactor vessel cooling (IVR-ERVC) as a severe accident management strategy. The previous researches focused on the quisi-steady state behavior of molten corium pool in the lower head and related in-vessel retention problem. However, questions of the feasibility of the in-vessel retention concept for high power density reactor and uncertainties due to layering effect require further studies. These researches are rather essential to consider the whole evolution of the accident including formation and growth of the molten pool and the characteristic of corium arrival in the lower head and molten pool behavior after the core debris remelting. The general objective of the LIVE program performed at FzK is to study the corium pool formation and behavior with emphasis on the transient behavior through the large scale 3-D experiments. In this report, description of LIVE experimental facility and results of performance test are briefly summarized and the process to select the simulant is depicted. Also, the results of LIVE L1 and L2 tests and analytical models are included. These experimental results are very useful to development and verification of the model of molten corium pool behavior

Advanced nondestructive examination of the reactor vessel head penetration tube welds

International Nuclear Information System (INIS)

Cvitanovic, M.; Zado, V.

1996-01-01

Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

Retention of mandibular advancement devices in the treatment of obstructive sleep apnea: an in vitro pilot study.

Science.gov (United States)

Vanderveken, Olivier M; Van de Heyning, Paul; Braem, Marc J

2014-05-01

In order for a mandibular advancement device (MAD) to be efficacious, it must remain seated on the teeth during sleep. Quantitative data on the retentive characteristics of MADs are currently unavailable. The present pilot study is the first to describe an in vitro setup testing the retentive characteristics of different monobloc MADs. A hydraulic cyclic test machine was used with MADs seated on dental casts to measure retention forces upon removal of the MADs. A custom-made monobloc (CM-mono), a thermoplastic monobloc (TP-mono), and a thermoplastic duobloc (TP-duo) configured as a monobloc were tested. Two protrusions were investigated, representing 25 and 65% of the maximal protrusion. The effects of the type of MAD, duration of the test, and amount of protrusion on removal forces were measured. The measured removal forces of all three MADs tested differed significantly, with the TP-duo showing the highest values (P tested. However, only the TP-duo showed a significant reduction in time in removal forces for both protrusion positions (P tested (P = 0.0074). This in vitro pilot study reveals significant differences in retention forces for the MADs tested. The findings are consistent with clinical effects of nightly loss of MADs as reported in the literature and are within the range of reported physiological mouth-opening forces. Future research is needed to determine the key design features of MADs that explain these differences.

Long Term Retention of Deuterium and Tritium in Alcator C-Mod

International Nuclear Information System (INIS)

FIORE, C.; LABOMBARD, B.; LIPSCHULTZ, B.; PITCHER, C.S.; SKINNER, C.H.; WAMPLER, WILLIAM R.

1999-01-01

We estimate the total in-vessel deuterium retention in Alcator C-Mod from a run campaign of about 1090 plasmas. The estimate is based on measurements of deuterium retained on 22 molybdenum tiles from the inner wall and divertor. The areal density of deuterium on the tiles was measured by nuclear reaction analysis. From these data, the in-vessel deuterium inventory is estimated to be about 0.1 gram, assuming the deuterium coverage is toroidally symmetric. Most of the retained deuterium is on the walls of the main plasma chamber, only about 2.5% of the deuterium is in the divertor. The D coverage is consistent with a layer saturated by implantation with ions and charge-exchange neutrals from the plasma. This contrasts with tokamaks with carbon plasma-facing components (PFC's) where long-term retention of tritium and deuterium is large and mainly in the divertor due to codeposition with carbon eroded by the plasma. The low deuterium retention in the C-Mod divertor is mainly due to the absence of carbon PFC's in C-Mod and the low erosion rate of Mo

Location and repair of air leaks in the ATF [Advanced Toroidal Facility] vacuum vessel

International Nuclear Information System (INIS)

Schwenterly, S.W.; Gabbard, W.A.; Schaich, C.R.; Yarber, J.L.

1989-01-01

On the basis of partial pressure rate-of-rise and base pressure measurements, it was determined that the Advanced Toroidal Facility (ATF) vacuum vessel had an air leak in the low 10 -4 mbar-ell/s range. Pinpointing this leak by conventional helium leak-checking procedures was not possible, because large portions of the outside of the vessel are covered by the helcial field coils and a structural shell. Various alternative leak-detection schemes that were considered are summarized and their advantages and disadvantages noted. In the method ultimately employed, gun-rubber patches of various sizes ranging from 12.7 by 12.7 cm to 20.3 by 30.5 cm were positioned on the inside surfaces of the vessel and evacuated by the leak detector (LD). After roughly 5% of the surface was inspected in this way, a leak of > 10 -5 mbar-ell/s was discovered and localized to an area of 5 by 5 cm. Dye penetrant applied to this area disclosed three pinholes. Two small slag pockets were discovered while these points were being ground out. After these were rewelded, no furthered leakage could be found in the repaired area. Global leak rates measured after the machine was reevacuated indicated that this leak was about 30% of the overall leak rate. 1 ref., 5 figs., 1 tab

Baking system for ports of experimental advanced super-conducting tokamak vacuum vessel and thermal stress analysis

International Nuclear Information System (INIS)

Cheng Yali; Bao Liman; Song Yuntao; Yao Damao

2006-01-01

The baking system of Experimental Advanced Super-Conducting Toakamk (EAST) vacuum vessel is necessary to obtain the baking temperature of 150 degree C. In order to define suitable alloy heaters and achieve their reasonable layouts, thermal analysis was carried out with ANSYS code. The analysis results indicate that the temperature distribution and thermal stress of most parts of EAST vacuum vessel ports are uniform, satisfied for the requirement, and are safe based on ASME criterion. Feasible idea on reducing the stress focus is also considered. (authors)

Synthetic analyses of the LAVA experimental results on in-vessel corium retention through gap cooling

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Cho, Young Ro; Koo, Kil Mo; Park, Rae Joon; Kim, Jong Hwan; Kim, Jong Tae; Ha, Kwang Sun; Kim, Sang Baik; Kim, Hee Dong

2001-03-01

LAVA(Lower-plenum Arrested Vessel Attack) has been performed to gather proof of gap formation between the debris and lower head vessel and to evaluate the effect of the gap formation on in-vessel cooling. Through the total of 12 tests, the analyses on the melt relocation process, gap formation and the thermal and mechanical behaviors of the vessel were performed. The thermal behaviors of the lower head vessel were affected by the formation of the fragmented particles and melt pool during the melt relocation process depending on mass and composition of melt and subcooling and depth of water. During the melt relocation process 10.0 to 20.0 % of the melt mass was fragmented and also 15.5 to 47.5 % of the thermal energy of the melt was transferred to water. The experimental results address the non-adherence of the debris to the lower head vessel and the consequent gap formation between the debris and the lower head vessel in case there was an internal pressure load across the vessel abreast with the thermal load induced by the thermite melt. The thermal behaviors of the lower head vessel during the cooldown period were mainly affected by the heat removal characteristics through this gap, which were determined by the possibilities of the water ingression into the gap depending on the melt composition of the corium simulant. The enhanced cooling capacity through the gap was distinguished in the Al 2 O 3 melt tests. It could be inferred from the analyses on the heat removal capacity through the gap that the lower head vessel could effectively cooldown via heat removal in the gap governed by counter current flow limits(CCFL) even if 2mm thick gap should form in the 30 kg Al 2 O 3 melt tests, which was also confirmed through the variations of the conduction heat flux in the vessel and rapid cool down of the vessel outer surface in the Al 2 O 3 melt tests. In the case of large melt mass of 70 kg Al 2 O 3 melt, however, the infinite possibility of heat removal through the

Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

International Nuclear Information System (INIS)

Maruyama, Yu; Yamano, Norihiro; Moriyama, Kiyofumi; Park, Hyun Sun; Kudo, Tamotsu; Yang, Yanhua; Sugimoto, Jun

1999-01-01

In-vessel debris coolability experiments were performed in ALPHA program at JAERI. Molten aluminum oxide (Al 2 O 3 ) was poured into a pool of water in a lower head experimental vessel. Post-test observation and measurement using an ultrasonic technique indicated the formation of the interfacial gap between the solidified Al 2 O 3 and the vessel wall. Thermal responses of the vessel wall implied that the interfacial gap acted initially as a thermal resistance and water subsequently penetrated into the interfacial gap. The maximum heat flux at the inner surface of the vessel facing to the solidified Al 2 O 3 was roughly evaluated to be ranged from 320 kW/m 2 to 600 kW/m 2 . A post-test analysis was conducted with CAMP code. The influence of the interfacial gap on thermal behavior of Al 2 O 3 and the vessel wall was examined. (authors)

Integral experiments on in-vessel coolability and vessel creep: results and analysis of the FOREVER-C1 test

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A. [Division of Nuclear Power Safety, Royal Institute of Technology, Drottning Kristinas Vaeg., Stockholm (Sweden)

1999-07-01

This paper describes the FOREVER (Failure Of REactor VEssel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The paper presents the experimental results and analysis of the first FOREVER-C1 test. During this experiment, the 1/10th scale pressure vessel, heated to about 900degC and pressurized to 26 bars, was subjected to creep deformation in a non-stop 24-hours test. The vessel wall displacement data clearly shows different stages of the vessel deformation due to thermal expansion, elastic, plastic and creep processes. The maximum displacement was observed at the lowermost region of the vessel lower plenum. Information on the FOREVER-C1 measured thermal characteristics and analysis of the observed thermal and structural behavior is presented. The coupled nature of thermal and mechanical processes, as well as the effect of other system conditions (such as depressurization) on the melt pool and vessel temperature responses are analyzed. (author)

Development of advanced design features for KNGR reactor vessel and internals

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new.

Development of advanced design features for KNGR reactor vessel and internals

International Nuclear Information System (INIS)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new

Simplified tools for measuring retention in care in antiretroviral treatment program in Ethiopia: cohort and current retention in care.

Science.gov (United States)

Assefa, Yibeltal; Worku, Alemayehu; Wouters, Edwin; Koole, Olivier; Haile Mariam, Damen; Van Damme, Wim

2012-01-01

Patient retention in care is a critical challenge for antiretroviral treatment programs. This is mainly because retention in care is related to adherence to treatment and patient survival. It is therefore imperative that health facilities and programs measure patient retention in care. However, the currently available tools, such as Kaplan Meier, for measuring retention in care have a lot of practical limitations. The objective of this study was to develop simplified tools for measuring retention in care. Retrospective cohort data were collected from patient registers in nine health facilities in Ethiopia. Retention in care was the primary outcome for the study. Tools were developed to measure "current retention" in care during a specific period of time for a specific "ART-age group" and "cohort retention" in care among patients who were followed for the last "Y" number of years on ART. "Probability of retention" based on the tool for "cohort retention" in care was compared with "probability of retention" based on Kaplan Meier. We found that the new tools enable to measure "current retention" and "cohort retention" in care. We also found that the tools were easy to use and did not require advanced statistical skills. Both "current retention" and "cohort retention" are lower among patients in the first two "ART-age groups" and "ART-age cohorts" than in subsequent "ART-age groups" and "ART-age cohorts". The "probability of retention" based on the new tools were found to be similar to the "probability of retention" based on Kaplan Meier. The simplified tools for "current retention" and "cohort retention" will enable practitioners and program managers to measure and monitor rates of retention in care easily and appropriately. We therefore recommend that health facilities and programs start to use these tools in their efforts to improve retention in care and patient outcomes.

Simplified tools for measuring retention in care in antiretroviral treatment program in Ethiopia: cohort and current retention in care.

Directory of Open Access Journals (Sweden)

Yibeltal Assefa

Full Text Available INTRODUCTION: Patient retention in care is a critical challenge for antiretroviral treatment programs. This is mainly because retention in care is related to adherence to treatment and patient survival. It is therefore imperative that health facilities and programs measure patient retention in care. However, the currently available tools, such as Kaplan Meier, for measuring retention in care have a lot of practical limitations. The objective of this study was to develop simplified tools for measuring retention in care. METHODS: Retrospective cohort data were collected from patient registers in nine health facilities in Ethiopia. Retention in care was the primary outcome for the study. Tools were developed to measure "current retention" in care during a specific period of time for a specific "ART-age group" and "cohort retention" in care among patients who were followed for the last "Y" number of years on ART. "Probability of retention" based on the tool for "cohort retention" in care was compared with "probability of retention" based on Kaplan Meier. RESULTS: We found that the new tools enable to measure "current retention" and "cohort retention" in care. We also found that the tools were easy to use and did not require advanced statistical skills. Both "current retention" and "cohort retention" are lower among patients in the first two "ART-age groups" and "ART-age cohorts" than in subsequent "ART-age groups" and "ART-age cohorts". The "probability of retention" based on the new tools were found to be similar to the "probability of retention" based on Kaplan Meier. CONCLUSION: The simplified tools for "current retention" and "cohort retention" will enable practitioners and program managers to measure and monitor rates of retention in care easily and appropriately. We therefore recommend that health facilities and programs start to use these tools in their efforts to improve retention in care and patient outcomes.

Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

International Nuclear Information System (INIS)

Yoon, Ho Jun; Suh, Kune Y.

1998-01-01

In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krsko

International Nuclear Information System (INIS)

Mihalina, M.; Spalj, S.; Glaser, B.

2016-01-01

The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krsko applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant's site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant's post-Fukushima improvements and adequacy with plant's procedures for severe accident mitigation. Conclusion is that NPP Krsko could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance's (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the

Advances in crack-arrest technology for reactor pressure vessels

International Nuclear Information System (INIS)

Bass, B.R.; Pugh, C.E.

1988-01-01

The Heavy-Section Steel Technology (HSST) Program at the Oak Ridge National Laboratory (ORNL) under the sponsorship of the US Nuclear Regulatory Commission is continuing to improve the understanding of conditions that govern the initiation, rapid propagation, arrest, and ductile tearing of cracks in reactor pressure vessel (RPV) steels. This paper describes recent advances in a coordinated effort being conducted under the HSST Program by ORNL and several subcontracting groups to develop the crack-arrest data base and the analytical tools required to construct inelastic dynamic fracture models for RPV steels. Large-scale tests are being carried out to generate crack-arrest toughness data at temperatures approaching and above the onset of Charpy upper-shelf behavior. Small- and intermediate-size specimens subjected to static and dynamic loading are being developed and tested to provide additional fracture data for RPV steels. Viscoplastic effects are being included in dynamic fracture models and computer programs and their utility validated through analyses of data from carefully controlled experiments. Recent studies are described that examine convergence problems associated with energy-based fracture parameters in viscoplastic-dynamic fracture applications. Alternative techniques that have potential for achieving convergent solutions for fracture parameters in the context of viscoplastic-dynamic models are discussed. 46 refs., 15 figs., 3 tabs

VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation)], E-mail: bechta@sbor.spb.su; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A. [Alexandrov Research Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V.; Almiashev, V.I. [Institute of Silicate Chemistry, Russian Academy of Sciences (ISCh RAS), St. Petersburg (Russian Federation); Lopukh, D.B. [SPb State Electrotechnical University (SPbGETU), St. Petersburg (Russian Federation); Bottomley, D. [EUROPAISCHE KOMMISSION, Joint Research Centre Institut fuer Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA/DEN/DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Rossendorf (FZR), Dresden (Germany); Fichot, F. [IRSN/DPAM/SEMCA, St. Paul lez Durance (France); Kymalainen, O. [FORTUM Nuclear Services Ltd., Espoo (Finland)

2009-06-15

The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere

International Nuclear Information System (INIS)

Bechta, S.V.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Vitol, S.A.; Sulatsky, A.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

2009-01-01

The long-term in-vessel corium retention (IVR) in the lower head bears a risk of the vessel wall deterioration caused by steel corrosion. The ISTC METCOR Project has studied physicochemical impact of prototypic coria having different compositions in air and steam and has generated valuable experimental data on vessel steel corrosion. It is found that the corrosion rate is sensitive to corium composition, but the composition of oxidizing above-melt atmosphere (air, steam) has practically no influence on it. A model of the corrosion process that integrates the experimental data, is proposed and used for development of correlations.

Trends in Tissue Engineering for Blood Vessels

Directory of Open Access Journals (Sweden)

Judee Grace Nemeno-Guanzon

2012-01-01

Full Text Available Over the years, cardiovascular diseases continue to increase and affect not only human health but also the economic stability worldwide. The advancement in tissue engineering is contributing a lot in dealing with this immediate need of alleviating human health. Blood vessel diseases are considered as major cardiovascular health problems. Although blood vessel transplantation is the most convenient treatment, it has been delimited due to scarcity of donors and the patient’s conditions. However, tissue-engineered blood vessels are promising alternatives as mode of treatment for blood vessel defects. The purpose of this paper is to show the importance of the advancement on biofabrication technology for treatment of soft tissue defects particularly for vascular tissues. This will also provide an overview and update on the current status of tissue reconstruction especially from autologous stem cells, scaffolds, and scaffold-free cellular transplantable constructs. The discussion of this paper will be focused on the historical view of cardiovascular tissue engineering and stem cell biology. The representative studies featured in this paper are limited within the last decade in order to trace the trend and evolution of techniques for blood vessel tissue engineering.

Complete resection of locally advanced ovarian carcinoma fixed to the pelvic sidewall and involving external and internal iliac vessels.

Science.gov (United States)

Nishikimi, Kyoko; Tate, Shinichi; Matsuoka, Ayumu; Shozu, Makio

2017-08-01

Locally advanced ovarian carcinomas may be fixed to the pelvic sidewall, and although these often involve the internal iliac vessels, they rarely involve the external iliac vessels. Such tumors are mostly considered inoperable. We present a surgical technique for complete resection of locally advanced ovarian carcinoma fixed to the pelvic sidewall and involving external and internal iliac vessels. A 69-year-old woman presented with ovarian carcinoma fixed to the right pelvic sidewall, which involved the right external and internal iliac arteries and veins and the right lower ureter, rectum, and vagina. We cut the external iliac artery and vein at the bifurcation and at the inguinal ligament to resect the external artery and vein. Then, we reconstructed the arterial and venous supplies of the right external artery and vein with grafts. After creating a wide space immediately inside of the sacral plexus to allow the tumor fixed to pelvic sidewall with the internal iliac vessels to move medially, we performed total internal iliac vessel resection. We achieved complete en bloc tumor resection with the right external and internal artery and vein, right ureter, vagina, and rectum adhering to the tumor. There were no intra- or postoperative complications, such as bleeding, graft occlusion, infection, or limb edema. Exfoliation from the sacral plexus and total resection with external and internal iliac vessels enables complete resection of the tumor fixed to the pelvic sidewall. Copyright © 2017 Elsevier Inc. All rights reserved.

Simulant melt experiments on performance of the in-vessel core catcher

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Sang-Baik; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

2007-01-01

In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an 'engineered gap' for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVA-GAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10 mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs

Corrosion of vessel steel during its interaction with molten corium

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Lopukh, D.B.; Gusarov, V.V.; Martinov, A.P.; Martinov, V.V.; Fieg, G.; Tromm, W.; Bottomley, D.; Tuomisto, H.

2006-01-01

This paper is concerned with corrosion of a cooled vessel steel structure interacting with molten corium in air and neutral (nitrogen) atmospheres during an in-vessel retention scenario. The data on corrosion kinetics at different temperatures on the heated steel surface, heat flux densities and oxygen potential in the system are presented. The post-test physico-chemical and metallographic analyses of melt samples and the corium-specimen ingot have clarified certain mechanisms of steel corrosion taking place during the in-vessel melt interaction

Study on natural circulation flow under reactor cavity flooding condition in advanced PWRs

International Nuclear Information System (INIS)

Tao Jun; Yang Jiang; Cao Jianhua; Lu Xianghui; Guo Dingqing

2015-01-01

Cavity flooding is an important severe accident management measure for the in-vessel retention of a degraded core by external reactor vessel cooling in advanced PWRs. A code simulation study on the natural circulation flow in the gap between the reactor vessel wall and insulation material under cavity flooding condition is performed by using a detailed mechanistic thermal-hydraulic code package RELAP 5. By simulating of an experiment carried out for studying the natural circulation flow for APR1400 shows that the code is applicable for analyzing the circulation flow under this condition. The analysis results show that heat removal capacity of the natural circulation flow in AP1000 is sufficient to prevent thermal failure of the reactor vessel under bounding heat load. Several conclusions can be drawn from the sensitivity analysis. Larger coolant inlet area induced larger natural circulation flow rate. The outlet should be large enough and should not be submerged by the cavity water to vent the steam-water mixture. In the implementation of cavity flooding, the flooding water level should be high enough to provide sufficient natural circulation driven force. (authors)

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)

Control Carbon to Prevent corium Stratification In-Vessel Retention

Energy Technology Data Exchange (ETDEWEB)

Go, A Ra; Hong, Seung Hyun; Kim, Sang Nyung [Kyung Hee Univ., Yongin (Korea, Republic of)

2013-10-15

As a result, the thermal margin decreases, and the nuclear reactor vessel may be destroyed. To control Carbons, which is the major cause of stratification, Ruthenium and Hafnium are inserted inside the lower reactor head which initiates a chemical reaction with Carbon. SPARTAN program is used to confirm a reaction probability which is measured in bond energy and strength etc. To analyze the possibility of bonding with Carbon, the initial property of Ruthenium and Carbon are measured during the calculated absorbing process. After following that theory, the Spartan program is able to determine if it can insert the metal. After verifying the combination of Ruthenium and Carbon, the Spartan program analyzes the impact of the Carbon to prevent the corium stratification. It determines the possibility of the success with the introduction of the IVR concept. Ruthenium is suitable to Carbon bonding process to decrease affect to corium behavior which do not form stratification. The metal which can combine with Carbon should be satisfied with temperature as high as 2800 .deg. C. Therefore, the further research works are determined by using the Spartan program to calculate the Carbon and Ruthenium bonding energy, and to check other bonding results as follows. After check the results, review this theory to insert the Ruthenium in reactor vessel. APR1400 and OPR1000, Korea Hydro and Nuclear power plant core meltdown accident has been evaluated a high level in severe accident. When the reactor core is melted down, it is stratified into the metal layer and the ceramic layer. As the heat conductivity of metal layer is higher than that of the ceramic layer, heat concentration occurs in the upper part of the bottom hemisphere which comes into contact with the metal layer.

Development of ball bearing in high temperature water for in-vessel type control rod drive mechanism of advanced marine reactor

Energy Technology Data Exchange (ETDEWEB)

Nunokawa, Hiroshi [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan); Yoritsune, Tsutomu; Imayoshi, Shou; Ochiai, Masa-aki; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kasahara, Yoshiyuki [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

2001-06-01

An advanced marine reactor MRX designed by Japan Atomic Energy Research Institute (JAERI) adopts an in-vessel type control rod drive mechanism, which is installed inside the reactor vessel. Since the in-vessel type control rod drive mechanism should work at a severe condition of a high temperature and high pressure water - 310degC and 12 MPa -, the JAERI has developed the components, a ball bearing of which especially is one of key technologies for realization of this type mechanism. The present report describes the development of the ball bearing containing a survey of materials, material screening tests on oxidation in an autoclave and rolling wear by a small facility, a trial fabrication of the full size ball bearing, and endurance test of it in the high temperature water. As a result, it was found from the development that the materials of cobalt alloy for both of the inner and outer races, cermet for the ball, and graphite for the retainer can satisfy the design condition of the ball bearing. (author)

Particle retention during long discharges in Tore Supra and JET

International Nuclear Information System (INIS)

Loarer, T.; Tsitrone, E.; Brosset, C.; Bucalossi, J.; Gunn, J.; Joffrin, E.; Monier-Garbet, P.; Pegourie, B.; Thomas, P.; Lomas, P.; Ongena, J.

2003-01-01

The particle balances and the associated particle retentions for the long discharge experiments performed in Tore-Supra and for the L and H mode discharges carried out in JET are reported in this paper. From the reported experiments, the same particle retention behaviors are observed in Tore-Supra and JET in spite of the differences between the plasma geometry and the confinement mode (respectively limiter L-mode and divertor H-mode). A particle retention up to 70-80% of Γ(puff) for the larger gas injection has been obtained in JET. The particle retention behavior observed with the gas puff appears to be strongly dominant in the particle retention process. Indeed, no influence has been noticed from the active pumping, the saturation of the recycling area (0.4 D/C), the precedent discharges history (in terms of total 'particles retained' in the vessel) and even from the disruptions (conditioning). Also, the outgassing between discharges becomes negligible in terms of particle recovering when Γ(puff) and/or the discharge duration are increased. Finally, neither the edge localized modes (ELMs type-I or III) nor the disruptions modify the reported behaviour. For ITER, the particle retention is strictly limited and from the presented results it seems that strong gas injection should be avoided. (A.C.)

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

Purpose: To prevent the retention of low temperature water and also prevent the thermal fatigue of the pressure vessel by making large the curvature radius of a pressure vessel of a feed water sparger fitting portion and accelerating the mixing of low-temperature water at the feed water sparger base and in-pile hot water. Constitution: The curvature radius of the corner of the feed water sparger fitting portion in a pressure vessel is formed largely. In-pile circulating water infiltrates up to the base portion of the feed water sparger to carry outside low-temperature water at the base part, which is mixed with in-pile hot water. Accordingly, low temperature water does not stay at the base portion of the feed water sparger and generation of thermal fatigue in the pressure vessel can be prevented and the safety of the BWR type reactor can be improved. (Yoshino, Y.)

Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

Science.gov (United States)

Kumar, B. Ramesh; Gangradey, R.

2012-11-01

Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

International Nuclear Information System (INIS)

Kumar, B Ramesh; Gangradey, R

2012-01-01

Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

Deuterium retention and desorption behavior in an advanced reduced-activation alloy

Energy Technology Data Exchange (ETDEWEB)

Noh, S.J., E-mail: sjnoh@dankook.ac.kr [Department of Applied Physics, Dankook University, Yongin-si, Gyeonggi-do 448-701 (Korea, Republic of); Kim, H.S.; Byeon, W.J.; Shin, H.W. [Department of Applied Physics, Dankook University, Yongin-si, Gyeonggi-do 448-701 (Korea, Republic of); Lee, Cheol Eui [Department of Physics, Korea University, Seoul 136-713 (Korea, Republic of); Lee, S.K. [Nuclear Fusion Development Division, Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

2017-07-15

We present the first experimental results of the deuterium retention and desorption behavior in an advanced reduced-activation alloy (ARAA) under development in Korea. For the in-situ measurement of desorbed gases from samples immediately after irradiation, a thermal desorption spectroscopy (TDS) system clustered with an inductively coupled plasma ion source has been built. Samples were and were not irradiated with helium ions at energies of 1.4, 3.5, and 5.0 keV and then continuously irradiated with 1.7-keV deuterium ions. TDS measurements were performed in situ immediately after deuterium irradiation and after exposure to air for one week. The amount of desorbed deuterium is the largest for the sample without helium irradiation from the TDS results measured in situ immediately after irradiation. Further, the amount of desorbed deuterium is significantly lowered when the helium energy is increased to 3.5 keV with no significant changes thereafter, indicating that the layer formed by implanted helium at near or deeper than the stopping range for 1.7-keV deuterium ions effectively acts as a barrier against deuterium diffusion into the depth. Because of the strong diffusivity of deuterium into the ambient atmosphere, the amounts of desorbed deuterium are greatly reduced for the samples without helium irradiation and with 1.4-keV helium irradiation after exposure to air for one week. In addition, our deuterium results for the ARAA are also compared with the results for F82H by other authors. - Highlights: •The first result of the deuterium retention and desorption in an ARAA is presented. •The ARAA was irradiated with helium and then continuously irradiated with deuterium. •TDS measurements were performed in situ immediately after deuterium irradiation. •TDS measurements were performed after exposure to air for one week. •The effects of helium irradiation and exposure to air were investigated.

Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

International Nuclear Information System (INIS)

Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R.

2008-03-01

This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to the

Ex-Vessel corium coolability and steam explosion energetics in nordic light water reactors

Energy Technology Data Exchange (ETDEWEB)

Dinh, T.N.; Ma, W.M.; Karbojian, A.; Kudinov, P.; Tran, C.T.; Hansson, C.R. [Royal Institute of Technology (KTH), (Sweden)

2008-03-15

This report presents advances and insights from the KTH's study on corium pool heat transfer in the BWR lower head; debris bed formation; steam explosion energetics; thermal hydraulics and coolability in bottom-fed and heterogeneous debris beds. Specifically, for analysis of heat transfer in a BWR lower plenum an advanced threedimensional simulation tool was developed and validated, using a so-called effective convectivity approach and Fluent code platform. An assessment of corium retention and coolability in the reactor pressure vessel (RPV) lower plenum by means of water supplied through the Control Rod Guide Tube (CRGT) cooling system was performed. Simulant material melt experiments were performed in an intermediate temperature range (1300-1600K) on DEFOR test facility to study formation of debris beds in high and low subcooled water pools characteristic of in-vessel and ex-vessel conditions. Results of the DEFOR-E scoping experiments and related analyses strongly suggest that porous beds formed in ex-vessel from a fragmented high-temperature debris is far from homogeneous. Calculation results of bed thermal hydraulics and dryout heat flux with a two-dimensional thermal-hydraulic code give the first basis to evaluate the extent by which macro and micro inhomogeneity can enhance the bed coolability. The development and validation of a model for two-phase natural circulation through a heated porous medium and its application to the coolability analysis of bottom-fed beds enables quantification of the significant effect of dryout heat flux enhancement (by a factor of 80-160%) due to bottom coolant injection. For a qualitative and quantitative understanding of steam explosion, the SHARP system and its image processing methodology were used to characterize the dynamics of a hot liquid (melt) drop fragmentation and the volatile liquid (coolant) vaporization. The experimental results provide a basis to suggest that the melt drop preconditioning is instrumental to

Transarterial chemoembolization through collateral vessels in hepatocellular carcinoma

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hye; Han, Joon Koo; Chung, Jin Wook; Park, Jae Hyung; Han, Man Chung [Seoul National University College of Medicine, Seoul (Korea, Republic of)

1993-11-15

We performed 70 procedures of transarterial chemoembolization (TAE) through extrahepatic collateral vessels (n=27) or parasitic feeders (n=18) in 45 hepatocellular carcinoma patients. The collaterals developed after interruption of the hepatic artery due to repeated TAE (n=17), surgical ligation (n=7)and primary celiac occlusion (n=3). Radiologic findings suggest the existence of parasitic or collateral supply for hepatocellular carcinoma were 1) a focal defect of Lipiodol retention on CT or plain film after TAE via the hepatic artery, 2) dilated and tortuous vessels around the mass on angiography, 3) persistent elevation of the level of serum alpha-fetoprotein or continuous clinical symptoms in spite of sufficient devascularization of the tumor via the hepatic artery, and 4) radiological findings of direct invasion into adjacent organ. The sites of the catheter placement were the inferior phrenic artery(n=19), omental branches(n=16), periportal collaterals (n=6), pancreaticodenal arcade (n=3), gastroduodenal artery(n=3), internal mammary artery (n=2), intercosal artery (n=2), lateral thoracic artery (n=1), bronchial artery (n=1), and colic branches (n=1). Masses feeded by the inferior phrenic and chest wall collaterals were usually located at the dome area of the liver, and the omental and gastroduodenal collaterals developed in the masses located at the inferior tip of the liver. After TAE via collateral vessels, 37 patients underwent follow-up study. In 18 cases(48%), the tumor favorably responded to TAE. Specific complications of collateral TAE were epigastric soreness (n=10), severe shoulder pain (n=4), and embolization of the spinal artery during embolization through the intercostal artery (n=1). In conclusion, various extrahepatic collateals are important alternative or addition routes for effective chemoembolization in patients with advanced hepatoma, and early recognition of the parasitic supply and the effort to perform TAE via collaterals is very

Transarterial chemoembolization through collateral vessels in hepatocellular carcinoma

International Nuclear Information System (INIS)

Kim, Ji Hye; Han, Joon Koo; Chung, Jin Wook; Park, Jae Hyung; Han, Man Chung

1993-01-01

We performed 70 procedures of transarterial chemoembolization (TAE) through extrahepatic collateral vessels (n=27) or parasitic feeders (n=18) in 45 hepatocellular carcinoma patients. The collaterals developed after interruption of the hepatic artery due to repeated TAE (n=17), surgical ligation (n=7)and primary celiac occlusion (n=3). Radiologic findings suggest the existence of parasitic or collateral supply for hepatocellular carcinoma were 1) a focal defect of Lipiodol retention on CT or plain film after TAE via the hepatic artery, 2) dilated and tortuous vessels around the mass on angiography, 3) persistent elevation of the level of serum alpha-fetoprotein or continuous clinical symptoms in spite of sufficient devascularization of the tumor via the hepatic artery, and 4) radiological findings of direct invasion into adjacent organ. The sites of the catheter placement were the inferior phrenic artery(n=19), omental branches(n=16), periportal collaterals (n=6), pancreaticodenal arcade (n=3), gastroduodenal artery(n=3), internal mammary artery (n=2), intercosal artery (n=2), lateral thoracic artery (n=1), bronchial artery (n=1), and colic branches (n=1). Masses feeded by the inferior phrenic and chest wall collaterals were usually located at the dome area of the liver, and the omental and gastroduodenal collaterals developed in the masses located at the inferior tip of the liver. After TAE via collateral vessels, 37 patients underwent follow-up study. In 18 cases(48%), the tumor favorably responded to TAE. Specific complications of collateral TAE were epigastric soreness (n=10), severe shoulder pain (n=4), and embolization of the spinal artery during embolization through the intercostal artery (n=1). In conclusion, various extrahepatic collateals are important alternative or addition routes for effective chemoembolization in patients with advanced hepatoma, and early recognition of the parasitic supply and the effort to perform TAE via collaterals is very

Reactor Vessel Surveillance Program for Advanced Reactor

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kyeong-Hoon; Kim, Tae-Wan; Lee, Gyu-Mahn; Kim, Jong-Wook; Park, Keun-Bae; Kim, Keung-Koo

2008-10-15

This report provides the design requirements of an integral type reactor vessel surveillance program for an integral type reactor in accordance with the requirements of Korean MEST (Ministry of Education, Science and Technology Development) Notice 2008-18. This report covers the requirements for the design of surveillance capsule assemblies including their test specimens, test block materials, handling tools, and monitors of the surveillance capsule neutron fluence and temperature. In addition, this report provides design requirements for the program for irradiation surveillance of reactor vessel materials, a layout of specimens and monitors in the surveillance capsule, procedures of installation and retrieval of the surveillance capsule assemblies, and the layout of the surveillance capsule assemblies in the reactor.

Assessment of reactor vessel integrity (ARVI)

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R. E-mail: sehgal@ne.kth.se; Theerthan, A.; Giri, A.; Karbojian, A.; Willschuetz, H.G.; Kymaelaeinen, O.; Vandroux, S.; Bonnet, J.M.; Seiler, J.M.; Ikkonen, K.; Sairanen, R.; Bhandari, S.; Buerger, M.; Buck, M.; Widmann, W.; Dienstbier, J.; Techy, Z.; Kostka, P.; Taubner, R.; Theofanous, T.; Dinh, T.N

2003-04-01

The cost-shared project ARVI (assessment of reactor vessel integrity) involves a total of nine organisations from Europe and USA. The objective of the ARVI Project is to resolve the safety issues that remain unresolved for the melt vessel interaction phase of the in-vessel progression of a severe accident. The work consists of experiments and analysis development. Four tests were performed in the EC-FOREVER Programme, in which failure was achieved in-vessels employing the French pressure vessel steel. The tests were analysed with the commercial code ANSYS-Multiphysics, and the codes SYSTUS+ and PASULA, and quite good agreement was achieved for the failure location. Natural convection experiments in stratified pools have been performed in the SIMECO and the COPO facilities, which showed that much greater heat is transferred downwards for immiscible layers or before layers mix. A model for gap cooling and a set of simplified models for the system codes have been developed. MVITA code calculations have been performed for the Czech and Hungarian VVERs, towards evaluation of the in-vessel melt retention accident management scheme. Tests have been performed at the ULPU facility with organised flow for vessel external cooling. Considerable enhancement of the critical heat flux (CHF) was obtained. The ARVI Project has reached the halfway stage. This paper presents the results obtained thus far from the project.

Coupled thermo-mechanical creep analysis for boiling water reactor pressure vessel lower head

International Nuclear Information System (INIS)

Villanueva, Walter; Tran, Chi-Thanh; Kudinov, Pavel

2012-01-01

Highlights: ► We consider a severe accident in a BWR with melt pool formation in the lower head. ► We study the influence of pool depth on vessel failure mode with creep analysis. ► There are two modes of failure; ballooning of vessel bottom and a localized creep. ► External vessel cooling can suppress creep and subsequently prevent vessel failure. - Abstract: In this paper we consider a hypothetical severe accident in a Nordic-type boiling water reactor (BWR) at the stage of relocation of molten core materials to the lower head and subsequent debris bed and then melt pool formation. Nordic BWRs rely on reactor cavity flooding as a means for ex-vessel melt coolability and ultimate termination of the accident progression. However, different modes of vessel failure may result in different regimes of melt release from the vessel, which determine initial conditions for melt coolant interaction and eventually coolability of the debris bed. The goal of this study is to define if retention of decay-heated melt inside the reactor pressure vessel is possible and investigate modes of the vessel wall failure otherwise. The mode of failure is contingent upon the ultimate mechanical strength of the vessel structures under given mechanical and thermal loads and applied cooling measures. The influence of pool depth and respective transient thermal loads on the reactor vessel failure mode is studied with coupled thermo-mechanical creep analysis. Efficacy of control rod guide tube (CRGT) cooling and external vessel wall cooling as potential severe accident management measures is investigated. First, only CRGT cooling is considered in simulations revealing two different modes of vessel failure: (i) a ‘ballooning’ of the vessel bottom and (ii) a ‘localized creep’ concentrated within the vicinity of the top surface of the melt pool. Second, possibility of in-vessel retention with CRGT and external vessel cooling is investigated. We found that the external vessel

Examination of VVER-1000 Reactor Pressure Vessel

International Nuclear Information System (INIS)

Matokovic, A.; Picek, E.; Markulin, K.

2008-01-01

The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

In-vessel core debris retention through external flooding of the reactor pressure vessel. SCDAP/RELAP5 assessment for the SBWR lower head

International Nuclear Information System (INIS)

Heel, A.M.J.M. van.

1995-09-01

In this report the results are discussed from various analyses on the feasibility and phenomenology of the External Flooding (EF) concept for an SBWR lower head, filled with a large heat generating corium mass. In applying External Flooding as an accident management strategy after or during core melt down, the lower drywell is filled with water up to a level where a large portion of the Reactor Pressure Vessel (RPV) is flooded. The purpose of this method is to establish cooling of the vessel wall, that is challenged by the heat load resulting from the corium, in such a way that its structural integrity if not endangered. The analysis discussed in this report focus on the thermal response of the vessel wall and the ex-vessel boiling processes under the conditions described above. For these analyses the SCDAP/REALP5 MOD 3.1 code was used. The major outcome of the calculations is, that a major part of the vessel wall remains well below themelting temperature of carbon steel, as long as flooding of the external surface of the lower head is established. The SCDAP/RELAP5 analyses indicated that low-quality Critical Heat Flux (CHF) was not exceeded, under all the conditions that had been tested. However, a comaprison of the heat fluxes, as calculated in RELAP5, with the CHF values obtained from the Zuber correlation and the Vishnev correction factor (for boiling at inclined surfaces) proved that CHF values, based on these criteria, were exceeded in several surface points of the lower head mesh. The correlations, as resident in the current version of RELAP5 MOD 3.1, might lead to over-estimation of CHF for the EF analyses discussed in this report. The use of the more conservative Zuber correlation with the Vishnev correction factor is recommended for EF analyses. (orig.)

Experimental studies of oxidic molten corium-vessel steel interaction

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V.

2001-01-01

The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere

Experimental studies of oxidic molten corium-vessel steel interaction

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. E-mail: niti-npc@sbor.net; Khabensky, V.B.; Vitol, S.A.; Krushinov, E.V.; Lopukh, D.B.; Petrov, Yu.B.; Petchenkov, A.Yu.; Kulagin, I.V.; Granovsky, V.S.; Kovtunova, S.V.; Martinov, V.V.; Gusarov, V.V

2001-12-01

The experimental results of molten corium-steel specimen interaction with molten corium on the 'Rasplav-2' test facility are presented. In the experiments, cooled vessel steel specimens positioned on the molten pool bottom and uncooled ones lowered into the molten pool were tested. Interaction processes were studied for different corium compositions, melt superheating and in alternative (inert and air) overlying atmosphere. Hypotheses were put forward explaining the observed phenomena and interaction mechanisms. The studies presented in the paper were aimed at the detection of different corium-steel interaction mechanisms. Therefore certain identified phenomena are more typical of the ex-vessel localization conditions than of the in-vessel corium retention. Primarily, this can be referred to the phenomena of low-temperature molten corium-vessel steel interaction in oxidizing atmosphere.

Pharmacological and physical vessel modulation strategies to improve EPR-mediated drug targeting to tumors.

Science.gov (United States)

Ojha, Tarun; Pathak, Vertika; Shi, Yang; Hennink, Wim E; Moonen, Chrit T W; Storm, Gert; Kiessling, Fabian; Lammers, Twan

2017-09-15

The performance of nanomedicine formulations depends on the Enhanced Permeability and Retention (EPR) effect. Prototypic nanomedicine-based drug delivery systems, such as liposomes, polymers and micelles, aim to exploit the EPR effect to accumulate at pathological sites, to thereby improve the balance between drug efficacy and toxicity. Thus far, however, tumor-targeted nanomedicines have not yet managed to achieve convincing therapeutic results, at least not in large cohorts of patients. This is likely mostly due to high inter- and intra-patient heterogeneity in EPR. Besides developing (imaging) biomarkers to monitor and predict EPR, another strategy to address this heterogeneity is the establishment of vessel modulation strategies to homogenize and improve EPR. Over the years, several pharmacological and physical co-treatments have been evaluated to improve EPR-mediated tumor targeting. These include pharmacological strategies, such as vessel permeabilization, normalization, disruption and promotion, as well as physical EPR enhancement via hyperthermia, radiotherapy, sonoporation and phototherapy. In the present manuscript, we summarize exemplary studies showing that pharmacological and physical vessel modulation strategies can be used to improve tumor-targeted drug delivery, and we discuss how these advanced combination regimens can be optimally employed to enhance the (pre-) clinical performance of tumor-targeted nanomedicines. Copyright © 2017 Elsevier B.V. All rights reserved.

Performances of helium, neon and argon glow discharges for reduction of fuel hydrogen retention in tungsten, stainless steel and graphite

International Nuclear Information System (INIS)

Hino, T.; Yamauchi, Y.; Kimura, Y.; Matsumoto, A.; Nishimura, K.; Ueda, Y.

2012-11-01

It is quite important to investigate the performance of glow discharge conditionings for controls of in-vessel tritium (T) inventory and hydrogen recycling. For this purpose, first, the deuterium (D) retentions in tungsten (W), graphite (C) and stainless steel (SS) were measured. The retention in W was not small as expected, several times larger than that of SS, although the retention in SS was one order smaller than that of C. Such the large retention in W is owing to the growth of rough surface structure produced by plasma irradiations. For reduction of deuterium retention in W, SS and C, second, inert gas (He, Ne, Ar) glow discharges were conducted under the same condition, and these performances were compared. The removal ratio of deuterium retention was highest in He discharge, and lowest in Ar discharge. These values are well explained by the numerical analyses using SRIM code. The removal ratios for SS and C were significantly large, but quite small for W. This reason is again owing to the rough surface structure in W. For W, thirdly, the hydrogen isotope exchange and the wall baking experiments were conducted. It is found that the wall backing with a temperature higher than 700 K can well reduce the retention, and the hydrogen isotope exchange using deuterium glow discharge is also useful to reduce the tritium retention in the wall. The present results significantly contribute to control the fuel hydrogen retention and to reduce the in-vessel tritium inventory in fusion reactors. (author)

An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

International Nuclear Information System (INIS)

Murphy, J.G.; Corradini, M.L.

1997-01-01

The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

Induction heating of a spherical aluminum moderator vessel for the Advanced Neutron Source (ANS)

International Nuclear Information System (INIS)

Yousuf, A.

1994-01-01

This task was to identify and design a heating system to apply 15 kW of heat to a cold source vessel to simulate the Advanced Neutron Source reactor. This research project aims at the analysis of the induction heating of a spherical aluminum moderator vessel. Computer modeling is presented for the design and analysis of the induction heating system. The objective is to apply 15 kW of heat as uniformly as possible to the outer wall of a 410 mm diameter sphere of thickness 1.5 mm. The report also aims at the analysis of a system model which is simulated using the Eddycuff electromagnetic software. The computer model is built with the finite element analysis software Patran. The induction heating system analysis shows that the predicted performance is in close agreement with the computer simulated data. Hardware constraints such as power supplies and matching load are also analyzed in terms of performance and cost. Physical modeling is also suggested, in which the coil and the workpiece are scaled down

Introduction to Test Facility for Iodine Retention in Filtered Containment Venting System

Energy Technology Data Exchange (ETDEWEB)

Jung, Jaehoon; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol [KAERI, Daejeon (Korea, Republic of)

2015-05-15

In many countries the implementation of FCVS's is under discussion to mitigate fission product release not only in the short-term but also in the long-term view. To verify the performance of FCVS, the large-scaled tests have been performed such as advanced containment experiments (ACE), the iodine and aerosol retention rate test facility (JAVA), etc. The elemental and organic iodides are the main gaseous iodine species in the containment atmosphere. For the iodine retention, experimental programs have confirmed the existence of gaseous organic iodine in some cases in higher concentrations than for gaseous molecular iodine (I{sub 2}). The Reaction of Methyl iodide (CH{sub 3}I) with surfaces and the removal by containment filters and scrubbers is less efficient in comparison to molecular iodine. In the recent years, an experimental and analytical work has been conducted at the Paul Scherrer Institute (PSI) to develop a process leading to a fast, comprehensive and reliable retention of volatile iodine species in aqueous solutions. New FCVS test facility to verify the performance of FCVS is designed and under construction. The iodine retention tests are planned with elemental iodine or with organic iodide loaded carrier gas consisting of pure non-condensable gas, pure steam and of typical mixtures of non-condensable gas/steam. This paper introduces the iodine generation and measurement system for the iodine retention test of FCVS. In severe accidents elemental and organic iodides are the main gaseous iodine species in the containment atmosphere. Release of the gaseous species in sufficient quantities from containment to environment generates a risk for public health. The filtered containment venting systems (FCVS) can considerably reduce the leakage of radioactive materials to the environment. New integral test facility is prepared to verify a performance of the FCVS. The test facility consists of a test vessel, thermal-hydraulic, and aerosol/iodine generation and

Baking results of KSTAR vacuum vessel

International Nuclear Information System (INIS)

Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M.

2009-01-01

The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported

Baking results of KSTAR vacuum vessel

Energy Technology Data Exchange (ETDEWEB)

Kim, S. T.; Kim, Y. J.; Kim, K. M.; Im, D. S.; Joung, N. Y.; Yang, H. L.; Kim, Y. S.; Kwon, M. [National Fusion Research Institute, Daejeon (Korea, Republic of)

2009-05-15

The Korea Superconducting Tokamak Advanced Research (KSTAR) is an advanced superconducting tokamak designed to establish a scientific and technological basis for an attractive fusion reactor. The fusion energy in the tokamak device is released through fusion reactions of light atoms such as deuterium or helium in hot plasma state, of which temperature reaches several hundreds of millions Celsius. The high temperature plasma is created in the vacuum vessel that provides ultra high vacuum status. Accordingly, it is most important for the vacuum condition to keep clean not only inner space but also surface of the vacuum vessel to make high quality plasma. There are two methods planned to clean the wall surface of the KSTAR vacuum vessel. One is surface baking and the other is glow discharge cleaning (GDC). To bake the vacuum vessel, De-Ionized (DI) water is heated to 130 .deg. C and circulated in the passage between double walls of the vacuum vessel (VV) in order to bake the surface. The GDC operation uses hydrogen and inert gas discharges. In this paper, general configuration and brief introduction of the baking result will be reported.

An Approach to Understanding Cohesive Slurry Settling, Mobilization, and Hydrogen Gas Retention in Pulsed Jet Mixed Vessels

Energy Technology Data Exchange (ETDEWEB)

Gauglitz, Phillip A.; Wells, Beric E.; Fort, James A.; Meyer, Perry A.

2009-05-22

The Hanford Waste Treatment and Immobilization Plant (WTP) is being designed and built to pretreat and vitrify a large portion of the waste in Hanford’s 177 underground waste storage tanks. Numerous process vessels will hold waste at various stages in the WTP. Some of these vessels have mixing-system requirements to maintain conditions where the accumulation of hydrogen gas stays below acceptable limits, and the mixing within the vessels is sufficient to release hydrogen gas under normal conditions and during off-normal events. Some of the WTP process streams are slurries of solid particles suspended in Newtonian fluids that behave as non-Newtonian slurries, such as Bingham yield-stress fluids. When these slurries are contained in the process vessels, the particles can settle and become progressively more concentrated toward the bottom of the vessels, depending on the effectiveness of the mixing system. One limiting behavior is a settled layer beneath a particle-free liquid layer. The settled layer, or any region with sufficiently high solids concentration, will exhibit non-Newtonian rheology where it is possible for the settled slurry to behave as a soft solid with a yield stress. In this report, these slurries are described as settling cohesive slurries.

In-vessel dust and tritium control strategy in ITER

Energy Technology Data Exchange (ETDEWEB)

Shimada, M., E-mail: michiya.shimada@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France); Pitts, R.A.; Ciattaglia, S.; Carpentier, S.; Choi, C.H.; Dell Orco, G.; Hirai, T.; Kukushkin, A.; Lisgo, S.; Palmer, J.; Shu, W.; Veshchev, E. [ITER Organization, Route de Vinon-sur-Verdon, 13115 St. Paul-lez-Durance (France)

2013-07-15

A baseline strategy for dust and tritium-inventory control and recovery in ITER has been established and preparations are underway for its implementation. Limits on dust and tritium-inventory are an integral part of the ITER safety case and are fixed at 1 kg for tritium, 1000 kg for mobilisable dust and 11 kg (beryllium)/76 kg (tungsten) for dust on hot surfaces. Maximum average T-retention rates of ∼1 g/shot are estimated for baseline inductive operation at Q{sub DT} = 10, suggesting that the in-vessel T-retention could reach the administrative limit of 640 g in as little as ∼2 months of operation. Baking is expected to remove a significant fraction of the T co-deposited on the divertor targets. Despite large uncertainties, dust quantities are expected to remain well below safety limits over the divertor cassette lifetime. In situ aspiration during divertor cassette exchange is foreseen as the main dust removal technique.

In-vessel tritium retention and removal in ITER

International Nuclear Information System (INIS)

Federici, G.; Anderl, R.A.

1998-01-01

The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

In-vessel tritium retention and removal in ITER

Energy Technology Data Exchange (ETDEWEB)

Federici, G. [ITER JWS Garching Co-Center (Germany); Anderl, R.A. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States). Idaho National Engineering and Environmental Lab.; Andrew, P. [JET Joint Undertaking, Abingdon (United Kingdom)] [and others

1998-06-01

The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the

Surgical tactic of high bilateral abdominal testicular retention in patient with a prune belly syndrome

Directory of Open Access Journals (Sweden)

I. A. Panchenko

2014-11-01

Full Text Available Bilateral abdominal cryptorchism in combination with other defects of urogenital system and a prune belly syndrome keep within a syndrome of connective tissue dysplasia. In our medical center developed the method of surgical correction of a high bilateral abdominal testicular retention with preservation of vessels.

Surgical tactic of high bilateral abdominal testicular retention in patient with a prune belly syndrome

Directory of Open Access Journals (Sweden)

I. A. Panchenko

2013-01-01

Full Text Available Bilateral abdominal cryptorchism in combination with other defects of urogenital system and a prune belly syndrome keep within a syndrome of connective tissue dysplasia. In our medical center developed the method of surgical correction of a high bilateral abdominal testicular retention with preservation of vessels.

Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

Direct vessel inclined injection system for reduction of emergency core coolant direct bypass in advanced reactors

International Nuclear Information System (INIS)

Yoon, Sang H.; Lee, Jong G.; Suh, Kune Y.

2006-01-01

Multidimensional thermal hydraulics in the APR1400 (Advanced Power Reactor 1400 MWe) downcomer during a large-break loss-of-coolant accident (LBLOCA) plays a pivotal role in determining the capability of the safety injection system. APR1400 adopts the direct vessel injection (DVI) method for more effective core penetration of the emergency core cooling (ECC) water than the cold leg injection (CLI) method in the OPR1000 (Optimized Power Reactor 1000 MWe). The DVI method turned out to be prone to occasionally lack in efficacious delivery of ECC to the reactor core during the reflood phase of a LBLOCA, however. This study intends to demonstrate a direct vessel inclined injection (DVII) method, one of various ideas with which to maximize the ECC core penetration and to minimize the direct bypass through the break during the reflood phase of a LBLOCA. The 1/7 scaled down THETA (Transient Hydrodynamics Engineering Test Apparatus) tests show that a vertical inclined nozzle angle of the DVII system increases the downward momentum of the injected ECC water by reducing the degree of impingement on the reactor downcomer, whereby lessening the extent of the direct bypass through the break. The proposed method may be combined with other innovative measures with which to ensure an enough thermal margin in the core during the course of a LBLOCA in APR1400

Light-water-reactor pressure-vessel surveillance dosimetry using solid-state track recorders

International Nuclear Information System (INIS)

Ruddy, F.H.; Roberts, J.H.; Gold, R.; Preston, C.C.

1983-07-01

The accumulation of neutron dose by the pressure vessel of an operating nuclear power plant results in damage in the form of steel embrittlement. In order to ascertain the safe operating lifetime of the reactor pressure vessel, dosimetric measurements must be made to evaluate the neutron dose to the pressure vessel and relate this dose to the cumulative radiation damage. Advanced dosimetry techniques are being evaluated for surveillance of operating reactors. Solid-state track recorder (SSTR) techniques are included among these advanced dosimetry techniques. Described herein are low neutron fluence calibration and standardization measurements that are being carried out in pressure vessel mockup benchmark neutron fields in the USA, Belgium, and England. In addition, high fluence SSTR dosimetry capsules have been irradiated with metallurgical specimens in a pressure vessel mockup facility. The design and deployment of advances SSTR dosimetry capsules in operating power reactors are also described

An experimental study of hypervapotron structure in external reactor vessel cooling

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

2016-07-15

Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

Experimental study on in-vessel debris coolability during severe accident

International Nuclear Information System (INIS)

Kim, S. B.; Park, R. J.; Kim, H. D.

2002-05-01

A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al 2 O 3 /Fe (or Al 2 O 3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

Modeling irradiation embrittlement in reactor pressure vessel steels

International Nuclear Information System (INIS)

Odette, G.R.

1998-01-01

As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 10, numerical modeling of irradiation embrittlement in reactor vessel steels are introduced. Physically-based models are developed and their role in advancing the state-of-the-art of predicting irradiation embrittlement of RPV steels is stressed

Retention characteristics of hydrogen isotopes in JT-60U

International Nuclear Information System (INIS)

Masaki, K.; Sugiyama, K.; Hayashi, T.; Ochiai, K.; Gotoh, Y.; Shibahara, T.; Hirohata, Y.; Oya, Y.; Miya, N.; Tanabe, T.

2005-01-01

Erosion/deposition distribution and hydrogen isotope behavior in the JT-60U plasma-facing wall were investigated. Distribution of the tritium, which was produced by D-D nuclear reaction, was not correlated with erosion/deposition distribution. The tritium distribution can be explained by the distribution of high-energy tritium ion-implantation due to ripple loss. Deuterium distribution in the divertor region was different from the tritium distribution and not well correlated with the deposition. The highest D/C was ∼0.05 at the bottom of the outer dome wing, which is much less than that observed in other tokamaks. For the deuterium retention, at least two retention processes (ion-implantation and co-deposition) were found on the dome region. The systematic dust collection gave the small amount of dust (∼7 g: 0.2 mg/s production) in the whole vessel of JT-60U. Deposition was observed at the remote area of the outer divertor region

Multiphase flow in ex-vessel coolability: development of an innovative concept

International Nuclear Information System (INIS)

Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific Advanced Light Water Reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability

Inter-vessels in-service inspection of Super-Phenix

International Nuclear Information System (INIS)

Asty, M.; Saglio, R.; Viard, J.; Lerat, B.

1984-01-01

The vessels design of fast breeder reactor Super-Phenix enables inspection during operating time. A self-moving machine -MIR- has been built up especially for that purpose. It is able to carry out visual and ultrasonorous inspection. MIR structure is that of a tetrahedron, all tops of which are fitted with two wheels, as for traction and direction. The wheels are leaning on booth the two vessels. Thanks to a computer-assisted control system, MIR is able to move along in every part of the inter-vessels space. Studies have been carried on at the French Commissariat a l'Energie Atomique, by two Sections of the advanced technologies Service. After outlining MIR working conditions, its main characteristics are described [fr

Intraabdominal laparoscopy-assisted "open" vessel ligation of testicular vessels: a potential treatment for varicocele.

Science.gov (United States)

Miyano, Go; Miyahara, Katsumi; Halibieke, Abudebieke; Lane, Geoffrey J; Okazaki, Tadaharu; Yamataka, Atsuyuki

2011-10-01

We tested our laparoscopy-assisted "open" ligation (LOL) technique on testicular vessels. We ligated the left testicular artery and vein (TAV) in 8-week-old male Wister rats using LOL (LOL group; n=10) or laparotomy (open group; n=10). In LOL, a 0-degree laparoscope was introduced through a 5-mm epigastric trocar. A 3-mm grasper was used to expose the left TAV. A lapa-her-closure (LHC) needle loaded with 3-0 SurgiPro was directly inserted into the left lower quadrant where the left TAV should be and advanced under the vessels, and the suture material was released leaving one end outside. The LHC was then withdrawn a little and advanced again over the vessels to grasp the end of the suture material just released to bring it outside. This was proximally repeated. The two ends of both sutures were conventionally tied outside, and the knot was passed through the insertion site and tightened around the vessels. In the open group, the left TAV were ligated using two 3-0 SurgiPro ties. In both groups, the right side was left intact. All rats were sacrificed 2 weeks postoperatively, and both testes were examined with hematoxylin and eosin. Treatment time was 5-7 minutes for LOL and 7-8 minutes for the open group. Postoperative recovery was uneventful. No adhesions were present between the ligated vessels and bowel in any rat. Histopathology of all left testes showed coagulative necrosis of germinal cells and seminiferous tubules; all right testes were normal. LOL appears to be as effective as open ligation and may find application for treating varicocele.

Optical Measurement Technologies for High Temperature, Radiation Exposure, and Corrosive Environments—Significant Activities and Findings: In-vessel Optical Measurements for Advanced SMRs

Energy Technology Data Exchange (ETDEWEB)

Anheier, Norman C.; Cannon, Bret D.; Qiao, Hong (Amy); Suter, Jonathan D.

2012-09-01

Development of advanced Small Modular Reactors (aSMRs) is key to providing the United States with a sustainable, economically viable, and carbon-neutral energy source. The aSMR designs have attractive economic factors that should compensate for the economies of scale that have driven development of large commercial nuclear power plants to date. For example, aSMRs can be manufactured at reduced capital costs in a factory and potentially shorter lead times and then be shipped to a site to provide power away from large grid systems. The integral, self-contained nature of aSMR designs is fundamentally different than conventional reactor designs. Future aSMR deployment will require new instrumentation and control (I&C) architectures to accommodate the integral design and withstand the extreme in-vessel environmental conditions. Operators will depend on sophisticated sensing and machine vision technologies that provide efficient human-machine interface for in-vessel telepresence, telerobotic control, and remote process operations. The future viability of aSMRs is dependent on understanding and overcoming the significant technical challenges involving in-vessel reactor sensing and monitoring under extreme temperatures, pressures, corrosive environments, and radiation fluxes

Accuracy and initial clinical experience with measurement software (advanced vessel analysis) in three-dimensional imaging

International Nuclear Information System (INIS)

Abe, Toshi; Hirohata, Masaru; Tanigawa, Hitoshi

2002-01-01

Recently, the clinical benefits of three dimensional (3D) imaging, such as 3D-CTA and 3D-DSA, in cerebro-vascular disease have been widely recognized. Software for quantitative analysis of vascular structure in 3D imaging (advanced vessel analysis: AVA) has been developed. We evaluated AVA with both phantom studies and a few clinical cases. In spiral and curvy aluminum tube phantom studies, the accuracy of diameter measurements was good in 3D images produced from data set generated by multi-detector row CT or rotational angiography. The measurement error was less than 0.03 mm on aluminum tube phantoms that were 3 mm and 5 mm in diameter. In the clinical studies, the differences of carotid artery diameter measurements between 2D-DSA and 3D-DSA was less than 0.3 mm in. The measurement of length, diameter and angle by AVA should provide useful information for planning surgical and endovascular treatments of cerebro-vascular disease. (author)

New sacrificial material for ex-vessel core catcher

Energy Technology Data Exchange (ETDEWEB)

Komlev, Andrei A., E-mail: komlev@kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Nuclear Power Safety Division, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Almjashev, Vyacheslav I., E-mail: vac@mail.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Bechta, Sevostian V., E-mail: bechta@safety.sci.kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Khabensky, Vladimir B., E-mail: vladimirkhabensky@gmail.com [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Granovsky, Vladimir S., E-mail: gran@niti.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Gusarov, Victor V., E-mail: victor.v.gusarov@gmail.com [Ioffe Institute, 26 Polytekhnicheskaya Str., St. Petersburg, 194021 (Russian Federation)

2015-12-15

A new functional (sacrificial) material has been developed in the Fe{sub 2}O{sub 3}–SrO–Al{sub 2}O{sub 3}–CaO system based on strontium hexaferrite ceramic in concrete matrix. The method of producing SM has been advanced technologically; this technological effectiveness allows the SM to be used in ex-vessel core catchers with corium spreading as well as in crucible-type core catchers. Critical properties regarding the efficiency of SM in ex-vessel core catchers, such as porosity, pycnometric density, apparent density, solidus and liquidus temperatures, and water content have been measured. Suitable fractions of SrFe{sub 12}O{sub 19} and high alumina cement (HAC) were found in the SM based on thermodynamic analysis of the SM/corium interaction. The use of sacrificial steel as an additional heat adsorption component in the core catcher allowed us to increase the mass fraction range of SrFe{sub 12}O{sub 19} in the SM from 0.3−0.5 to 0.3–0.85. The activation temperature of the SM/corium interaction has been shown to correspond to the liquidus temperature of the local composition at the SM/corium interface. The calculated value of this temperature was 1716 °C. Analysis of phase transformations in the SrO–Fe{sub 2}O{sub 3} system revealed advantages of the SrFe{sub 12}O{sub 19}–based sacrificial material compared with the Fe{sub 2}O{sub 3}-contained material owing to the time proximity of SrFe{sub 12}O{sub 19} decomposition and corium interaction activation. - Highlights: • A sacrificial material (SM) was developed for ex-vessel core catcher. • Suitable proportions in the SrFe{sub 12}O{sub 19}–Al{sub 2}O{sub 3}·CaO–Fe system were determined. • Hydrogen release limitation was shown for ex-vessel corium retention with the SM. • Calculated temperature of the active initiation of corium/SM interaction is 1716 °C. • Functional properties of the SM were measured.

Simulant Basis for the Standard High Solids Vessel Design

Energy Technology Data Exchange (ETDEWEB)

Peterson, Reid A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Suffield, Sarah R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Daniel, Richard C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gauglitz, Phillip A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wells, Beric E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-09-01

This document provides the requirements for a test simulant suitable for demonstrating the mixing requirements for the Single High Solids Vessel Design (SHSVD). This simulant has not been evaluated for other purposes such as gas retention and release or erosion. The objective of this work is to provide an underpinning for the simulant properties based on actual waste characterization.

Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

International Nuclear Information System (INIS)

Ishino, Shiori

2007-01-01

The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

Efficiency of thermal outgassing for tritium retention measurement and removal in ITER

Directory of Open Access Journals (Sweden)

G. De Temmerman

2017-08-01

Full Text Available As a licensed nuclear facility, ITER must limit the in-vessel tritium (T retention to reduce the risks of potential release during accidents, the inventory limit being set at 1kg. Simulations and extrapolations from existing experiments indicate that T-retention in ITER will mainly be driven by co-deposition with beryllium (Be eroded from the first wall, with co-deposits forming mainly in the divertor region but also possibly on the first wall itself. A pulsed Laser-Induced Desorption (LID system, called Tritium Monitor, is being designed to locally measure the T-retention in co-deposits forming on the inner divertor baffle of ITER. Regarding tritium removal, the baseline strategy is to perform baking of the plasma-facing components, at 513K for the FW and 623K for the divertor. Both baking and laser desorption rely on the thermal desorption of tritium from the surface, the efficiency of which remains unclear for thick (and possibly impure co-deposits. This contribution reports on the results of TMAP7 studies of this efficiency for ITER-relevant deposits.

Thiazolidinediones and Edema: Recent Advances in the Pathogenesis of Thiazolidinediones-Induced Renal Sodium Retention.

Science.gov (United States)

Horita, Shoko; Nakamura, Motonobu; Satoh, Nobuhiko; Suzuki, Masashi; Seki, George

2015-01-01

Thiazolidinediones (TZDs) are one of the major classes of antidiabetic drugs that are used widely. TZDs improve insulin resistance by activating peroxisome proliferator-activated receptor gamma (PPARγ) and ameliorate diabetic and other nephropathies, at least, in experimental animals. However, TZDs have side effects, such as edema, congestive heart failure, and bone fracture, and may increase bladder cancer risk. Edema and heart failure, which both probably originate from renal sodium retention, are of great importance because these side effects make it difficult to continue the use of TZDs. However, the pathogenesis of edema remains a matter of controversy. Initially, upregulation of the epithelial sodium channel (ENaC) in the collecting ducts by TZDs was thought to be the primary cause of edema. However, the results of other studies do not support this view. Recent data suggest the involvement of transporters in the proximal tubule, such as sodium-bicarbonate cotransporter and sodium-proton exchanger. Other studies have suggested that sodium-potassium-chloride cotransporter 2 in the thick ascending limb of Henle and aquaporins are also possible targets for TZDs. This paper will discuss the recent advances in the pathogenesis of TZD-induced sodium reabsorption in the renal tubules and edema.

Results of Large-Scale Testing on Effects of Anti-Foam Agent on Gas Retention and Release

Energy Technology Data Exchange (ETDEWEB)

Stewart, Charles W.; Guzman-Leong, Consuelo E.; Arm, Stuart T.; Butcher, Mark G.; Golovich, Elizabeth C.; Jagoda, Lynette K.; Park, Walter R.; Slaugh, Ryan W.; Su, Yin-Fong; Wend, Christopher F.; Mahoney, Lenna A.; Alzheimer, James M.; Bailey, Jeffrey A.; Cooley, Scott K.; Hurley, David E.; Johnson, Christian D.; Reid, Larry D.; Smith, Harry D.; Wells, Beric E.; Yokuda, Satoru T.

2008-01-03

The U.S. Department of Energy (DOE) Office of River Protection’s Waste Treatment Plant (WTP) will process and treat radioactive waste that is stored in tanks at the Hanford Site. The waste treatment process in the pretreatment facility will mix both Newtonian and non-Newtonian slurries in large process tanks. Process vessels mixing non-Newtonian slurries will use pulse jet mixers (PJMs), air sparging, and recirculation pumps. An anti-foam agent (AFA) will be added to the process streams to prevent surface foaming, but may also increase gas holdup and retention within the slurry. The work described in this report addresses gas retention and release in simulants with AFA through testing and analytical studies. Gas holdup and release tests were conducted in a 1/4-scale replica of the lag storage vessel operated in the Pacific Northwest National Laboratory (PNNL) Applied Process Engineering Laboratory using a kaolin/bentonite clay and AZ-101 HLW chemical simulant with non-Newtonian rheological properties representative of actual waste slurries. Additional tests were performed in a small-scale mixing vessel in the PNNL Physical Sciences Building using liquids and slurries representing major components of typical WTP waste streams. Analytical studies were directed at discovering how the effect of AFA might depend on gas composition and predicting the effect of AFA on gas retention and release in the full-scale plant, including the effects of mass transfer to the sparge air. The work at PNNL was part of a larger program that included tests conducted at Savannah River National Laboratory (SRNL) that is being reported separately. SRNL conducted gas holdup tests in a small-scale mixing vessel using the AZ-101 high-level waste (HLW) chemical simulant to investigate the effects of different AFAs, their components, and of adding noble metals. Full-scale, single-sparger mass transfer tests were also conducted at SRNL in water and AZ-101 HLW simulant to provide data for PNNL�

Problems in Pressure Vessel Design and Manufacture

Energy Technology Data Exchange (ETDEWEB)

Hellstroem, O [Uddeholms AB, Degerfors (Sweden); Nilson, Ragnar [AB Atomenergi, Nykoeping (Sweden)

1963-05-15

The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels.

Problems in Pressure Vessel Design and Manufacture

International Nuclear Information System (INIS)

Hellstroem, O.; Nilson, Ragnar

1963-05-01

The general desire by the power reactor process makers to increase power rating and their efforts to involve more advanced thermal behaviour and fuel handling facilities within the reactor vessels are accompanied by an increase in both pressure vessel dimensions and various difficulties in giving practical solutions of design materials and fabrication problems. In any section of this report it is emphasized that difficulties and problems already met with will meet again in the future vessels but then in modified forms and in many cases more pertinent than before. As for the increase in geometrical size it can be postulated that with use of better materials and adjusted fabrication methods the size problems can be taken proper care of. It seems likely that vessels of sufficient large diameter and height for the largest power output, which is judged as interesting in the next ten year period, can be built without developing totally new site fabrication technique. It is, however, supposed that such a fabrication technique will be feasible though at higher specific costs for the same quality requirements as obtained in shop fabrication. By the postulated use of more efficient vessel material with principally the same good features of easy fabrication in different stages such as preparation, welding, heat treatment etc as ordinary or slightly modified carbon steels the increase in wall thickness might be kept low. There exists, however, a development work to be done for low-alloy steels to prove their justified use in large reactor pressure vessels

Manufacture of EAST VS In-Vessel Coil

International Nuclear Information System (INIS)

Long, Feng; Wu, Yu; Du, Shijun; Jin, Huan; Yu, Min; Han, Qiyang; Wan, Jiansheng; Liu, Bin; Qiao, Jingchun; Liu, Xiaochuan; Li, Chang; Cai, Denggang; Tong, Yunhua

2013-01-01

Highlights: • ITER like Stainless Steel Mineral Insulation Conductor (SSMIC) used for EAST Tokamak VS In-Vessel Coil manufacture first time. • Research on SSMIC fabrication was introduced in detail. • Two sets totally four single-turn VS coils were manufactured and installed in place symmetrically above and below the mid-plane in the vacuum vessel of EAST. • The manufacture and inspection of the EAST VS coil especially the joint for the SSMIC connection was described in detail. • The insulation resistances of all the VS coils have no significant reduction after endurance test. -- Abstract: In the ongoing latest update round of EAST (Experimental Advanced Superconducting Tokamak), two sets of two single-turn Vertical Stabilization (VS) coils were manufactured and installed symmetrically above and below the mid-plane in the vacuum vessel of EAST. The Stainless Steel Mineral Insulated Conductor (SSMIC) developed for ITER In-Vessel Coils (IVCs) in Institute of Plasma Physics, Chinese Academy of Science (ASIPP) was used for the EAST VS coils manufacture. Each turn poloidal field VS coil includes three internal joints in the vacuum vessel. The middle joint connects two pieces of conductor which together form an R2.3 m arc segment inside the vacuum vessel. The other two joints connect the arc segment with the two feeders near the port along the toroidal direction to bear lower electromagnetic loads during operation. Main processes and tests include material performances checking, conductor fabrication, joint connection and testing, coil forming, insulation performances measurement were described herein

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Estimation of the contribution of gaps to tritium retention in the divertor of ITER

International Nuclear Information System (INIS)

Matveev, D; Kirschner, A; Litnovsky, A; Borodin, D; Samm, U; Schmid, K; Komm, M; Van Oost, G

2014-01-01

An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s −1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas. (paper)

Deuterium retention and surface modification of tungsten macrobrush samples exposed in FTU Tokamak

Science.gov (United States)

Maddaluno, G.; Giacomi, G.; Rufoloni, A.; Verdini, L.

2007-06-01

The effect of discrete structures such as macrobrush or castellated surfaces on power handling and deuterium retention of plasma facing components is to be assessed since such geometrical configurations are needed for increasing the lifetime of the armour to heat-sink joint. Four small macrobrush W and W + 1%La2O3 samples have been exposed in the Frascati Tokamak Upgrade (FTU) scrape-off layer up to the last closed flux surface by means of the Sample Introduction System. FTU is an all metal machine with no carbon source inside vacuum vessel; it exhibits ITER relevant energy and particle fluxes on the plasma facing components. Here, results on morphological surface changes (SEM), chemical composition (EDX) and deuterium retention (TDS) are reported.

Safety and Efficacy of Advanced Bipolar Vessel Sealing in Vaginal Hysterectomy: 1000 Cases.

Science.gov (United States)

Clavé, Henri; Clavé, Arnaud

2017-02-01

To determine the safety and efficacy of advanced electrothermal bipolar vessel sealing (EBVS) during vaginal hysterectomy by evaluating urinary complications, overall complications, and reoperation rate. A retrospective cohort (Canadian Task Force classification III). High-volume gynecologic surgeon practice, private hospital. One thousand consecutive patients who have undergone vaginal hysterectomy for benign conditions carried out with EBVS between January 2002 and December 2012. Vaginal hysterectomy performed using an EBVS device. One thousand consecutive patients underwent vaginal hysterectomy with advanced EBVS between January 2002 and December 2012 with an average age of 51.4 ± 8.9 years (range, 31-88) and mean weight and body mass index of 57.4 ± 7.2 kg (range, 42-105) and 25.8 ± 4.2 kg/m 2 (range, 19.1-38.9), respectively. Eighty-five percent of patients (852/1000) were healthy without any severe systemic disease. A single experienced surgeon performed all vaginal hysterectomies with EBVS, specifically by not applying traction during thermofusion to avoid hemorrhage, amputating the cervix to transform the uterus to an apple shape to facilitate a vaginal approach and rotation of the uterus, and placing bi-clamp forceps on the edge of the uterus and not at a 45- or 90-degree angle. Wound closure was completed with a continuous suture. Eleven urinary complications (1.1%) were recorded (10 bladder mechanical injuries and 1 vesicovaginal fistula). This was not statistically different from the rate of .64% previously reported in the FINHYST study (p = .15). The overall rate of complications was 5.3%, and 20 patients (2.0%) required reoperation. The presence of uterine scar tissue (odds ratio, 5.5; 95% confidence interval, 1.6-19.2) and larger uterus size (odds ratio, 2.5; 95% confidence interval, 1.01-19.2) were associated with a higher risk of urinary complications. The use of EBVS during vaginal hysterectomy results in urinary and overall

In-vessel coolability and retention of a core melt

International Nuclear Information System (INIS)

Theofanous, T.G.; Liu, C.; Additon, S.

1997-01-01

The efficacy of external flooding of a reactor vessel as a severe accident management strategy is assessed for an AP600-like reactor design. The overall approach is based on the Risk Oriented Accident Analysis Methodology (ROAAM), and the assessment includes consideration of bounding scenarios and sensitivity studies, as well as arbitrary parametric evaluations that allow the delineation of the failure boundaries. The technical treatment in this assessment includes: (a) new data on energy flow from either volumetrically heated pools or non-heated layers on top, boiling and critical heat flux in inverted, curved geometries, emissivity of molten (superheated) samples of steel, and chemical reactivity proof tests, (b) a simple but accurate mathematical formulation that allows prediction of thermal loads by means of convenient hand calculations, (c) a detailed model programmed on the computer to sample input parameters over the uncertainty ranges, and to produce probability distributions of thermal loads and margins for departure from nucleate boiling at each angular position on the lower head, and (d) detailed structural evaluations that demonstrate that departure from nucleate boiling is a necessary and sufficient criterion for failure. Quantification of the input parameters is carried out for an AP600-like design, and the results of the assessment demonstrate that lower head failure is open-quotes physically unreasonable.close quotes Use of this conclusion for any specific application is subject to verifying the required reliability of the depressurization and cavity-flooding systems, and to showing the appropriateness (in relation to the database presented here, or by further testing as necessary) of the thermal insulation design and of the external surface properties of the lower head, including any applicable coatings

Registered nurse retention strategies in nursing homes: a two-factor perspective.

Science.gov (United States)

Hunt, Selina R; Probst, Janice C; Haddock, Kathlyn S; Moran, Robert; Baker, Samuel L; Anderson, Ruth A; Corazzini, Kirsten

2012-01-01

As the American population ages and the proportion of individuals over the age of 65 expands, the demand for high-quality nursing home care will increase. However, nursing workforce instability threatens care quality and sustainability in this sector. Despite increasing attention to nursing home staff turnover, far less is known about registered nurse (RN) retention. In this study, the relationships between retention strategies, employee benefits, features of the practice environment, and RN retention were explored. Further, the utility of Herzberg's two-factor theory of motivation as a framework for nursing home retention studies was evaluated. This study was a secondary analysis of the nationally representative 2004 National Nursing Home Survey. The final sample of 1,174 participating nursing homes were either certified by Medicare or Medicaid or licensed by state agencies. We used a weighted multinomial logistic regression using an incremental approach to model the relationships. Although most nursing homes offered some combination of retention programs, the majority of strategies did not have a significant association with the level of RN retention reported by facilities. Director of nursing tenure and other extrinsic factors had the strongest association with RN retention in adjusted analyses. To improve RN retention, organizations may benefit greatly from stabilizing nursing home leadership, especially the director of nursing position. Second, managers of facilities with poor retention may consider adding career ladders for advancement, awarding attendance, and improving employee benefits. As a behavioral outcome of motivation and satisfaction, retention was not explained as expected using Herzberg's two-factor theory.

Recovery of resources for advanced life support space applications: effect of retention time on biodegradation of two crop residues in a fed-batch, continuous stirred tank reactor

Science.gov (United States)

Strayer, R. F.; Finger, B. W.; Alazraki, M. P.; Cook, K.; Garland, J. L.

2002-01-01

Bioreactor retention time is a key process variable that will influence costs that are relevant to long distance space travel or long duration space habitation. However. little is known about the effects of this parameter on the microbiological treatment options that are being proposed for Advanced Life Support (ALS) systems. Two bioreactor studies were designed to examine this variable. In the first one, six retention times ranging from 1.3 to 21.3 days--were run in duplicate, 81 working-volume continuous stirred tank reactors (CSTR) that were fed ALS wheat residues. Ash-free dry weight loss, carbon mineralization, soluble TOC reduction, changes in fiber content (cellulose, hemicellulose, and lignin), bacterial numbers, and mineral recoveries were monitored. At short retention times--1.33 days--biodegradation was poor (total: 16-20%, cellulose - 12%, hemicellulose - 28%) but soluble TOC was decreased by 75-80% and recovery of major crop inorganic nutrients was adequate, except for phosphorus. A high proportion of the total bacteria (ca. 83%) was actively respiring. At the longest retention time tested, 21.3 days, biodegradation was good (total: 55-60%, cellulose ca. 70%, hemicellulose - ca. 55%) and soluble TOC was decreased by 80%. Recovery of major nutrients, except phosphorus, remained adequate. A very low proportion of total bacteria was actively respiring (ca. 16%). The second bioreactor study used potato residue to determine if even shorter retention times could be used (range 0.25-2.0 days). Although overall biodegradation deteriorated, the degradation of soluble TOC continued to be ca. 75%. We conclude that if the goal of ALS bioprocessing is maximal degradation of crop residues, including cellulose, then retention times of 10 days or longer will be needed. If the goal is to provide inorganic nutrients with the smallest volume/weight bioreactor possible, then a retention time of 1 day (or less) is sufficient.

Thiazolidinediones and Edema: Recent Advances in the Pathogenesis of Thiazolidinediones-Induced Renal Sodium Retention

Directory of Open Access Journals (Sweden)

Shoko Horita

2015-01-01

Full Text Available Thiazolidinediones (TZDs are one of the major classes of antidiabetic drugs that are used widely. TZDs improve insulin resistance by activating peroxisome proliferator-activated receptor gamma (PPARγ and ameliorate diabetic and other nephropathies, at least, in experimental animals. However, TZDs have side effects, such as edema, congestive heart failure, and bone fracture, and may increase bladder cancer risk. Edema and heart failure, which both probably originate from renal sodium retention, are of great importance because these side effects make it difficult to continue the use of TZDs. However, the pathogenesis of edema remains a matter of controversy. Initially, upregulation of the epithelial sodium channel (ENaC in the collecting ducts by TZDs was thought to be the primary cause of edema. However, the results of other studies do not support this view. Recent data suggest the involvement of transporters in the proximal tubule, such as sodium-bicarbonate cotransporter and sodium-proton exchanger. Other studies have suggested that sodium-potassium-chloride cotransporter 2 in the thick ascending limb of Henle and aquaporins are also possible targets for TZDs. This paper will discuss the recent advances in the pathogenesis of TZD-induced sodium reabsorption in the renal tubules and edema.

76 FR 30374 - Cruise Vessel Safety and Security Act of 2010, Available Technology

Science.gov (United States)

2011-05-25

... http://www.regulations.gov on or before July 25, 2011 or reach the Docket Management Facility by that... that occur on the vessel, and provide law enforcement officials in the course and scope of an... industry best practices for placement and retention of video recording devices exist? If yes, please...

Carbon transport and fuel retention in JT-60U with high temperature operation based on postmortem analysis

Energy Technology Data Exchange (ETDEWEB)

Yoshida, M., E-mail: yoshida.masafumi@jaea.go.jp [Japan Atomic Energy Agency, Mukoyama 801-1, Naka-shi, Ibaraki-ken 311-0193 (Japan); Tanabe, T.; Adachi, A. [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Hayashi, T.; Nakano, T.; Fukumoto, M.; Yagyu, J.; Miyo, Y.; Masaki, K.; Itami, K. [Japan Atomic Energy Agency, Mukoyama 801-1, Naka-shi, Ibaraki-ken 311-0193 (Japan)

2013-07-15

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 10{sup 20} H + D s{sup −1}, and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world.

Carbon transport and fuel retention in JT-60U with high temperature operation based on postmortem analysis

International Nuclear Information System (INIS)

Yoshida, M.; Tanabe, T.; Adachi, A.; Hayashi, T.; Nakano, T.; Fukumoto, M.; Yagyu, J.; Miyo, Y.; Masaki, K.; Itami, K.

2013-01-01

Fuel retention rates and carbon re-deposition rates in the plasma shadowed areas, or tile gaps and remote areas, in JT-60U were measured. The total fuel retention rate in the plasma shadowed areas was more than two times higher than that in the carbon re-deposited layers on the plasma facing surfaces, or the divertor tiles. This is because of lower temperature in the plasma shadowed areas than in the plasma facing surfaces, which leads to high hydrogen saturation concentration, although the amount of the carbon re-deposited on the plasma shadowed areas was only 60% of that on the plasma facing surfaces. The total fuel retention rate in JT-60U, including previously determined for all the plasma facing areas, was evaluated to be 1.3 × 10 20 H + D s −1 , and this retention rate was lower than that in the other devices, due probably to high baking temperature operation in JT-60U. Distributions of the fuel retention and the carbon re-deposition in the whole in-vessel of a large tokamak were determined for the first time in the world

Novel Use of the GuideLiner Catheter to Deliver Rotational Atherectomy Burrs in Tortuous Vessels

Directory of Open Access Journals (Sweden)

Minh Vo

2014-01-01

Full Text Available Rotational atherectomy (RA for heavily calcified lesions is essential for improved stent delivery and stent expansion. In tortuous vessels it is often difficult to advance the burr without rotation and possible injury to the endothelium of healthy vessel. The GuideLiner catheter, a child in mother catheter, has recently been used to allow for increased support for delivery of stents through tortuous vessels. We report a novel use of the GuideLiner for the delivery of an RA burr in tortuous vessels requiring increased guide support.

The effectiveness of ERC advanced life support (ALS) provider courses for the retention of ALS knowledge.

Science.gov (United States)

Fischer, Henrik; Strunk, Guido; Neuhold, Stephanie; Kiblböck, Daniel; Trimmel, Helmut; Baubin, Michael; Domanovits, Hans; Maurer, Claudia; Greif, Robert

2012-02-01

Out-of-hospital emergency physicians in Austria need mandatory emergency physician training, followed by biennial refresher courses. Currently, both standardized ERC advanced life support (ALS) provider courses and conventional refresher courses are offered. This study aimed to compare the retention of ALS-knowledge of out-of-hospital emergency physicians depending on whether they had or had not participated in an ERC-ALS provider course since 2005. Participants (n=807) from 19 refresher courses for out-of-hospital emergency physicians answered eight multiple-choice questions (MCQ) about ALS based on the 2005 ERC guidelines. The pass score was 75% correct answers. A multivariate logistic regression analyzed differences in passing scores between those who had previously participated in an ERC-ALS provider course and those who had not. Age, gender, regularity of working as an out-of-hospital emergency physician and the self-reported number of real resuscitation efforts within the last 6months were entered as control variables. Out-of-hospital emergency physicians who had previously attended an ERC-ALS provider course had a significantly higher chance of passing the MCQ test (OR=1.60, p=0.015). Younger age (OR=0.95, pERC-ALS provider course since 2005 had a higher retention of ALS knowledge compared to non-ERC-ALS course participants. Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04

International Nuclear Information System (INIS)

2004-01-01

The 2004 International Congress on Advances in Nuclear Power Plants (ICAPP'04) provides a forum for the industry to exchange the latest ideas and research findings on nuclear plants from all perspectives. This conference builds on the success of last year's meeting held in Cordoba, Spain, and on the 2002 inaugural meeting held in Hollywood, Florida. Because of the hard work of many volunteers from around the world, ICAPP'04 has been successful in achieving its goal. More than 325 invited and contributed papers/presentations are part of this ICAPP. There are 5 invited plenary sessions and 70 technical sessions with contributed papers. The ICAPP'04 Proceedings contain almost 275 papers prepared by authors from 25 countries covering topics related to advances in nuclear power plant technology. The program by technical track deals with: 1 - Water-Cooled Reactor Programs and Issues (Status of All New Water-Cooled Reactor Programs; Advanced PWRs: Developmental Stage I; Advanced PWRs: Developmental Stage II; Advanced PWRs: Basic Design Stage; Advanced BWRs; Economics, Regulation, Licensing, and Construction; AP1000); 2 - High Temperature Gas Cooled Reactors (Pebble Bed Modular Reactors; Very High Temperature Reactors; HTR Fuels and Materials; Innovative HTRs and Fuel Cycles); 3 - Long Term Reactor Programs and Strategies (Supercritical Pressure Water Reactors; Lead-Alloy Fast Reactors; Sodium and Gas Fast Reactors; Status of Advanced Reactor Programs; Non-classical Reactor Concepts); 4 - Operation, Performance, and Reliability Management (Information Technology Effect on Plant Operation; Operation, Maintenance and Reliability; Improving Performance and Reducing O and M Costs; Plant Modernization and Retrofits); 5 - Plant Safety Assessment and Regulatory Issues (LOCA and non-LOCA Analysis Methodologies; LOCA and non-LOCA Plant Analyses; In-Vessel Retention; Containment Performance and Hydrogen Control; Advances in Severe Accident Analysis; Advances in Severe Accident

Trajectory planning of tokamak flexible in-vessel inspection robot

Energy Technology Data Exchange (ETDEWEB)

Wang, Hesheng [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Chen, Weidong, E-mail: wdchen@sjtu.edu.cn [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China); Lai, Yinping; He, Tao [Department of Automation, Shanghai Jiao Tong University, 200240 Shanghai (China); Key Laboratory of System Control and Information Processing, Ministry of Education of China, 200240 Shanghai (China)

2015-10-15

Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

Trajectory planning of tokamak flexible in-vessel inspection robot

International Nuclear Information System (INIS)

Wang, Hesheng; Chen, Weidong; Lai, Yinping; He, Tao

2015-01-01

Highlights: • A tokamak flexible in-vessel inspection robot is designed. • Two trajectory planning methods are used to ensure the full coverage of the first wall scanning. • The method is tested on a simulated platform of EAST with the flexible in-vessel inspection robot. • Experimental results show the effectiveness of the proposed algorithm. - Abstract: Tokamak flexible in-vessel inspection robot is mainly designed to carry a camera for close observation of the first wall of the vacuum vessel, which is essential for the maintenance of the future tokamak reactor without breaking the working condition of the vacuum vessel. A tokamak flexible in-vessel inspection robot is designed. In order to improve efficiency of the remote maintenance, it is necessary to design a corresponding trajectory planning algorithm to complete the automatic full coverage scanning of the complex tokamak cavity. Two different trajectory planning methods, RS (rough scanning) and FS (fine scanning), according to different demands of the task, are used to ensure the full coverage of the first wall scanning. To quickly locate the damage position, the first trajectory planning method is targeted for quick and wide-ranging scan of the tokamak D-shaped section, and the second one is for careful observation. Furthermore, both of the two different trajectory planning methods can ensure the full coverage of the first wall scanning with an optimal end posture. The method is tested on a simulated platform of EAST (Experimental Advanced Superconducting Tokamak) with the flexible in-vessel inspection robot, and the results show the effectiveness of the proposed algorithm.

Experimental Study of Interactions Between Sub-oxidized Corium and Reactor Vessel Steel

International Nuclear Information System (INIS)

Bechta, S.V.; Khabensky, V.B.; Granovsky, V.S.; Krushinov, E.V.; Vitol, S.A.; Gusarov, V.V.; Almiashev, V.I.; Lopukh, D.B.; Tromm, W.; Miassoedov, A.; Bottomley, D.; Fischer, M.; Piluso, P.; Altstadt, E.; Willschutz, H.G.; Fichoti, F.

2006-01-01

One of the critical factors in the analysis of in-vessel melt retention is the vessel strength. It is, in particular, sensitive to the thickness of intact vessel wall, which, in its turn, depends on the thermal conditions and physicochemical interactions with corium. Physicochemical interaction of prototypic UO 2 -ZrO 2 -Zr corium melt and VVER vessel steel was examined during the 2. Phase of the ISTC METCOR Project. Rasplav-3 test facility was used for conducting four tests, in which the Zr oxidation degree and interaction front temperature were varied; in one of the tests, stainless steel was added to the melt. Direct experimental measurements and post-test analyses were used for determining corrosion kinetics and maximum corrosion depth (i.e. the physicochemical impact of corium on the cooled vessel steel specimens), as well as the steel temperature conditions during the interaction, and finally the structure and composition of crystallized ingots, including the interaction zone. The minimum temperature on the interaction front boundary, which determined its final position and maximum corrosion depth was ∼ 1090 deg. C. An empirical correlation for calculation of corrosion kinetics has been derived. (authors)

Development of gap measurement technique in-vessel corium retention using ultrasonic pulse echo method

International Nuclear Information System (INIS)

Koo, Kil Mo; Kim, Jong Hwan; Kang, Kyung Ho; Kim, Sang Baik; Sim, Cheul Muu

1999-03-01

A gap between a molten material and a lower vessel is formed in the LAVA experiment, a phase 1 study of Sonata-IV program. In this technical report, quantitative results of the gap measurement using an off-line ultrasonic pulse echo method are presented. This report aims at development of an appropriate ultrasonics test method, by analyzing the problems from the external environmental reason and the internal characteristic reason. The signal analyzing methods to improve the S/N ratio in these problems are divided into the time variant synthesized signal analyzing method and the time invariant synthesized signal analyzing method. In this report, the possibility of the application of these two methods to the gap signal and the noise is considered. In this test, the signal of the propagational direction and reflectional direction through solid-liquid-solid specimen was analyzed to understand the behavior of the reflectional signal in a multi-layered structure by filling the gap with water between the melt and the lower head vessel. The quantitative gap measurement using the off-line ultrasonic pulse echo method was available for a little of the scanned region. But furtherly using DSP technique and imaging technique, the better results will be obtained. Some of the measured signals are presented as 2-dimensional spherical mapping method using distance and amplitude. Other signals difficult in quantitative measurement are saved for a new signal processing method. (author). 11 refs., 4 tabs., 54 figs

Antineutrophil cytoplasmic autoantibody-associated small-vessel vasculitis

NARCIS (Netherlands)

Kallenberg, Cees G. M.

Purpose of reviews This review focuses on recent advance in the diagnosis pathogenesis and treatment of antineutrophil cytoplasmic autoantibody-associated small-vessel vasculitis. Recent findings Antineutrophil cytoplasmic autoantibodies are closely associated with Wegener's granulomatosis and

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

OpenAIRE

KO, DO-YOUNG; KIM, KYU-HYUNG

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

Enhancement of downward-facing saturated boiling heat transfer by the cold spray technique

Energy Technology Data Exchange (ETDEWEB)

Sohag, Frauk A.; Beck, Faith R.; Mohanta, Lokanath; Cheung, Fan Bill; Segall, Albert E.; Eden, Timothy J.; Potter, John K. [Pennsylvania State University, University Park (United States)

2017-02-15

In-vessel retention by passive external reactor vessel cooling under severe accident conditions is a viable approach for retention of radioactive core melt within the reactor vessel. In this study, a new and versatile coating technique known as 'cold spray' that can readily be applied to operating and advanced reactors was developed to form a microporous coating on the outer surface of a simulated reactor lower head. Quenching experiments were performed under simulated in-vessel retention by passive external reactor vessel cooling conditions using test vessels with and without cold spray coatings. Quantitative measurements show that for all angular locations on the vessel outer surface, the local critical heat flux (CHF) values for the coated vessel were consistently higher than the corresponding CHF values for the bare vessel. However, it was also observed for both coated and uncoated surfaces that the local rate of boiling and local CHF limit vary appreciably along the outer surface of the test vessel. Nonetheless, results of this intriguing study clearly show that the use of cold spray coatings could enhance the local CHF limit for downward-facing boiling by > 88%.

Grade Retention at Primary School: What the Teachers in Zimbabwe Say

Directory of Open Access Journals (Sweden)

Muchemwa Stella

2017-03-01

Full Text Available Primary education is the base of all advanced education levels, developments and innovations; it should therefore be meticulously done so as not to disadvantage the learners. In a bid to do it well, those involved in primary education have tried and are still trying strategies that can give the learners maximum advantage, for instance, grade retention and grade promotion. This research aimed at assessing the Zimbabwean primary school teachers’ stance on grade retention. The researcher took the advantage of Block-Release students at Solusi University in April 2016 (these are teachers from all over Zimbabwe and other countries in Africa who come for their degree studies at Solusi University during the holidays to collect data through interviews from a purposive sample of these teachers. Collected data was coded and analyzed descriptively. Findings showed that: 56% of the respondents indicated that retention is taking place at the primary schools where they are teaching but usually on a minimal scale since it is usually done upon parents’ request; 67% of the respondents said that retention is necessary and helpful; retention can be most suitably done at infant grades (1-3; however, the effect of retention on pupils’ performance is not clear due to other factors that affect performance. Findings also showed that the Zimbabwean Government is neither for retention nor grade promotion but for ERI (Early Reading Initiative and PLAP (Performance Lag Address Programme which have been recently introduced.

Design features of the KSTAR in-vessel control coils

Energy Technology Data Exchange (ETDEWEB)

Kim, H.K. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)], E-mail: hkkim@nfri.re.kr; Yang, H.L.; Kim, G.H.; Kim, Jin-Yong; Jhang, Hogun; Bak, J.S.; Lee, G.S. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)

2009-06-15

In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm x 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

ACR-1000 design provisions for severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-700 (ACR-700) as an evolutionary advancement of the current CANDU 6 reactor. As a further advancement of the ACR design, AECL is currently developing the ACR-1000 for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving enhanced safety features, shorter construction schedule, high plant capacity factor, improved operations and maintenance, and increased operating life. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. The ACR-1000 design meets Canadian regulatory requirements and follows established international practice with respect to severe accident prevention and mitigation. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. While maintaining existing structures of CANDU reactors that provide inherent prevention and retention of core debris, the ACR-1000 design includes additional features for prevention and mitigation of severe accidents. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel is designed for debris retention by minimizing penetrations at the bottom periphery and by accommodating thermal and weight loads of the core debris. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes

Welding distortion control in double walled KSTAR vacuum vessel fabrication

International Nuclear Information System (INIS)

Oh, D. W.; Lee, G. T.; Kim, H. K.; Yang, H. L.; Bak, J. S.

2004-01-01

The KSTAR(Korea Superconducting Tokamak Advanced Research) vacuum vessel is designed to be a double walled structure made of 12mm thick 316LN stainless steel with a D shaped cross-section about 4 m height. Vacuum vessel was pre-fabricated in two parts, 180 degree and 157.5 degree sectors in toroidal direction to meet the transportation purpose. These two parts have to be welded on site with ±2mm allowable fabrication tolerances. 1/3 scaled mock-up model was used to estimate the welding distortion and to ensure the weld quality of vacuum vessel. Gas Tungsten Arc Welding(GTAW), which has been approved by procedure qualification test, was used during mock-up test and vacuum vessel site fabrication. Welding distortion could be managed by allowing for distortion in opposite direction, by applying high restraint using lots of strong backs, by controlling the welding heat input with symmetrical welding sequence. The integrity of the site welding joint was assured by radiographic test, ultrasonic test and leak test with helium detecting method

Limiting Factors for External Reactor Vessel Cooling

International Nuclear Information System (INIS)

Cheung, F.B.

2005-01-01

The method of external reactor vessel cooling (ERVC) that involves flooding of the reactor cavity during a severe accident has been considered a viable means for in-vessel retention (IVR). For high-power reactors, however, there are some limiting factors that might adversely affect the feasibility of using ERVC as a means for IVR. In this paper, the key limiting factors for ERVC have been identified and critically discussed. These factors include the choking limit for steam venting (CLSV) through the bottleneck of the vessel/insulation structure, the critical heat flux (CHF) for downward-facing boiling on the vessel outer surface, and the two-phase flow instabilities in the natural circulation loop within the flooded cavity. To enhance ERVC, it is necessary to eliminate or relax these limiting factors. Accordingly, methods to enhance ERVC and thus improve margins for IVR have been proposed and demonstrated, using the APR1400 as an example. The strategy is based on using two distinctly different methods to enhance ERVC. One involves the use of an enhanced vessel/insulation design to facilitate steam venting through the bottleneck of the annular channel. The other involves the use of an appropriate vessel coating to promote downward-facing boiling. It is found that the use of an enhanced vessel/insulation design with bottleneck enlargement could greatly facilitate the process of steam venting through the bottleneck region as well as streamline the resulting two-phase motions in the annular channel. By selecting a suitable enhanced vessel/insulation design, not only the CLSV but also the CHF limits could be significantly increased. In addition, the problem associated with two-phase flow instabilities and flow-induced mechanical vibration could be minimized. It is also found that the use of vessel coatings made of microporous metallic layers could greatly facilitate downward-facing boiling on the vessel outer surface. With vessel coatings, the local CHF limits at

ACR-1000: Enhanced response to severe accidents

International Nuclear Information System (INIS)

Popov, N.K.; Santamaura, P.; Shapiro, H.; Snell, V.G.

2006-01-01

Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

Performance experiments on the in-vessel core catcher during severe accidents

International Nuclear Information System (INIS)

Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Rho; Kim, Sang Baik

2004-01-01

A US-Korean International Nuclear Energy Research Initiative (INERI) project has been initiated by the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) to determine if IVR is feasible for high power reactors up to 1500 MWe by investigating the performance of enhanced ERVC and in-vessel core catcher. This program is initially focusing on the Korean Advanced Power Reactor 1400 MWe (APR1400) design. As for the enhancement of the coolability through the ERVC, boiling tests are conducted by using appropriate coating material on the vessel outer surface to promote downward facing boiling and selecting an improved vessel/insulation design to facilitate water flow and steam venting through the insulation in this program. Another approach for successful IVR are investigated by applying the in-vessel core catcher to provide an 'engineered gap' between the relocated core materials and the water-filled reactor vessel and a preliminary design for an in-vessel core catcher was developed during the first year of this program. Feasibility experiments using the LAVA facility, named LAVA-GAP experiments, are in progress to investigate the core catcher performance based on the conceptual design of the in-vessel core catcher proposed in this INERI project. The experiments were performed using 60kg of Al 2 O 3 thermite melt as a core material simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The hemispherical in-vessel core catcher was installed inside the lower head vessel maintaining a uniform gap of 10mm from the inner surface of the lower head vessel. Two types of the core catchers were used in these experiments. The first one was a single layered in-vessel core catcher without internal coating and the second one was a two layered in-vessel core catcher with an internal coating of 0.5mm-thick ZrO 2 via the plasma

Gender Differences in Academic Medicine: Retention, Rank, and Leadership Comparisons From the National Faculty Survey.

Science.gov (United States)

Carr, Phyllis L; Raj, Anita; Kaplan, Samantha E; Terrin, Norma; Breeze, Janis L; Freund, Karen M

2018-01-30

Prior studies have found that women in academic medicine do not advance or remain in their careers in parity with men. The authors examined a national cohort of faculty from the 1995 National Faculty Survey to identify predictors of advancement, retention, and leadership for women faculty. The authors followed 1,273 faculty at 24 medical schools in the continental United States for 17 years to identify predictors of advancement, retention, and leadership for women faculty. Schools were balanced for public or private status and the four Association of American Medical Colleges geographic regions. The authors used regression models to adjust for covariates: seniority, department, academic setting, and race/ethnicity. After adjusting for significant covariates women were less likely than men to achieve the rank of professor (OR = 0.57; 95% CI, 0.43-0.78) or to remain in academic careers (OR = 0.68; 95% CI, 0.49-0.94). When number of refereed publications were added to the model, differences by gender in retention and attainment of senior rank were no longer significant. Male faculty were more likely to hold senior leadership positions after adjusting for publications (OR = 0.49; 95% CI, 0.35-0.69). Gender disparities in rank, retention, and leadership remain across the career trajectories of the faculty cohort in this study. Women were less likely to attain senior-level positions than men, even after adjusting for publication-related productivity. Institutions must examine the climate for women to ensure their academic capital is fully utilized and equal opportunity exists for leadership.

Quantitative estimation of 123I-MIBG scintigraphy in neuroblastoma. Usefulness of the total body retention rate

International Nuclear Information System (INIS)

Okuyama, Chio; Ushijima, Yo; Sugihara, Hiroki; Nishimura, Tunehiko

2000-01-01

A new method of easily and simply quantifying 123 I-MIBG accumulation as a criterion for curative effect of chemotherapy in infants with neuroblastoma was developed. This method uses the data from two images: an early image (at 5-7.5 hr) and a delayed image (at 25-32 hr). Twenty infants with untreated neuroblastoma which showed clear accumulation of 123 I-MIBG at the primary site were examined. The differences between the accumulation counts on the early image and the delayed image showed that washout of 123 I-MIBG in the neuroblastomas was delayed in tumor regions. This finding indicated that the total body 123 I-MIBG retention rate reflects the total volume of the neuroblastoma. The 123 I-MIBG retention rate was significantly higher in patients with advanced stage neuroblastoma with systemic metastases, and there was a good correlation between the retention rate and tumor markers (VMA and HVA values in urine). The response to chemotherapy paralleled the change in markers. These results suggested that the total body 123 I-MIBG retention rate is useful as a criterion for curative effect in advanced neuroblastoma. (K.H.)

Integration of ITER in-vessel diagnostic components in the vacuum vessel

International Nuclear Information System (INIS)

Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

2009-01-01

The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

Plasma wall interaction and tritium retention in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Amarescu, E.; Ascione, G.

1996-01-01

The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced an impressive number of record breaking results including core fusion power, ∼ 2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently, the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. The authors present recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues

Plasma wall interaction and tritium retention in TFTR

International Nuclear Information System (INIS)

Skinner, C.H.; Amarescu, E.; Ascione, G.

1997-01-01

The Tokamak Fusion Test Reactor (TFTR) has been operating safely and routinely with deuterium-tritium fuel for more than two years. In this time, TFTR has produced a number of record breaking results including core fusion power, ∝2 MW/m 3 , comparable to that expected for ITER. Advances in wall conditioning via lithium pellet injection have played an essential role in achieving these results. Deuterium-tritium operation has also provided a special opportunity to address the issues of tritium recycling and retention. Tritium retention over two years of operation was approximately 40%. Recently the in-torus tritium inventory was reduced by half through a combination of glow discharge cleaning, moist-air soaks, and plasma discharge cleaning. The tritium inventory is not a constraint in continued operations. Recent results from TFTR in the context of plasma wall interactions and deuterium-tritium issues are presented. (orig.)

CAP vessel monitoring. Programme, measurement and neutron calculation

International Nuclear Information System (INIS)

Farrugia, J.M.; Nimal, J.C.; Totth, B.; Lloret, R.; Perdreau, R.

1982-03-01

Starting with the design of the CAP (Prototype Advanced NSSS), a programme for pressure vessel monitoring has been prepared, including dosimetry. The dosimetry programme encompasses activation dosimeters (Cu, Nb, Co) and fission dosimeters ( 237 Np, 238 U) installed either inside the pressure vessel with the monitoring test-samples, or in a counting tube outside the pressure vessel. In the first place, a description of the method for neutronic calculation is given; such calculations use the codes ANISN and MERCURE 4 allowing assessment of the neutron spectrum seen by the detectors and the related reaction coefficient. This is followed by a description of the instrumentation. The initial dosimetry results available after the initial operating cycles concur with calculations [fr

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Recruitment and Early Retention of Women with Advanced Breast Cancer in a Complementary and Alternative Medicine Trial

Directory of Open Access Journals (Sweden)

Alla Sikorskii

2011-01-01

Full Text Available More than 80% of women with breast cancer are now reported to be using complementary and alternative medicine (CAM therapies during conventional treatment. A randomized clinical trial (RCT of reflexology with late stage breast cancer patients serves as the data source for this article. The purposes were to investigate: (i reasons for refusal to participate in a RCT of reflexology; (ii the differences between those who completed the baseline interview and those who dropped out before baseline; and (iii the utility of the Palliative Prognostic Score (PPS as a prognostic screening tool in minimizing early attrition (before baseline from the trial. Eligible women (N = 400 approached at 12 cancer centers in the Midwest had advanced breast cancer, were on chemotherapy or hormonal therapy, and had a PPS of 11 or less. Comparisons of those who dropped out early (N = 33 to those who stayed in the trial (N = 240 were carried out using Wilcoxon rank, t-, chi-squared and Fisher's exact tests. The reasons of being “too sick” or “overwhelmed” were given by less than 12% of the women who refused to participate. There was a higher early dropout rate among black women compared to other (primarily white women (P = .01. Cancer recurrence and metastasis, age, and the PPS were not predictive of early retention of women. Specialized techniques may be needed to ensure black women remain in the trial once consented. Women with advanced disease were likely to enter and remain in the trial despite deterioration in health.

Analysis of materials in connection with corium melt retention in WWER reactor vessels. Final report for the period 15 October 1995 - 14 October 1996

International Nuclear Information System (INIS)

Efanov, A.D.

1997-02-01

Analysis of the state of severe accident codes being developed in Russia describing processes of corium - reactor pressure vessel interaction during severe accidents showed that at present there is no reliable validated and verified code. This study considered some of the most advanced severe accident codes which include models of heat generating liquid convections and RPV tolerance capacity however the possibility of physico-chemical interactions is not being considered. The final report demonstrates one of the examples for the settlement modelling of processes of reactor core debris cooling and corium pool mirror cooling with the use of the KOSTER 2 Code developed at IPPE. It was shown that cooling by water spray within some definite region of drop dimensions (0.5 / 4 mm in diameter) could provide the opportunity of the WWER type RPV integrity preservation under the accepted conditions. Due to some uncertainties in calculation modules obtained results are recommended to be treated as preliminary. (author). 26 refs, figs, tabs

Tumor angiogenesis in advanced stage ovarian carcinoma.

Science.gov (United States)

Hollingsworth, H C; Kohn, E C; Steinberg, S M; Rothenberg, M L; Merino, M J

1995-07-01

Tumor angiogenesis has been found to have prognostic significance in many tumor types for predicting an increased risk of metastasis. We assessed tumor vascularity in 43 cases of advanced stage (International Federation of Gynecologists and Obstetricians stages III and IV) ovarian cancer by using the highly specific endothelial cell marker CD34. Microvessel counts and stage were associated with disease-free survival and with overall survival by Kaplan-Meier analysis. The plots show that higher stage, higher average vessel count at 200x (200x avg) and 400x (400x avg) magnification and highest vessel count at 400x (400x high) magnification confer a worse prognosis for disease-free survival. Average vessel count of less than 16 (400x avg, P2 = 0.01) and less than 45 (200x avg, P2 = 0.026) suggested a better survival. Similarly, a high vessel count of less than 20 (400x high, P2 = 0.019) conferred a better survival as well. The plots suggest that higher stage, higher average vessel count at 200x and 400x, and highest vessel count at 200x and 400x show a trend to worse overall survival as well. With the Cox proportional hazards model, stage was the best predictor of overall survival, however, the average microvessel count at 400x was found to be the best predictor of disease-free survival. These results suggest that analysis of neovascularization in advanced stage ovarian cancer may be a useful prognostic factor.

77 FR 15973 - Fisheries Off West Coast States; Highly Migratory Species Fisheries; Swordfish Retention Limits

Science.gov (United States)

2012-03-19

... Magnuson-Stevens Fishery Conservation and Management Act (MSA) to modify retention limits for swordfish... (RFA). Based on the average observed annual catches by this vessel and the market values for that catch... due to this comment. The second critical comment, submitted on behalf of the Turtle Island Restoration...

Urinary retention in women.

Science.gov (United States)

Juma, Saad

2014-07-01

This review is a summary of the most pertinent published studies in the literature in the last 18 months that address cause, diagnosis, and management of urinary retention in women. Symptoms, uroflow, and pressure-flow studies have a low predictive value for and do not correlate with elevated postvoid residual urine (PVR). Anterior and posterior colporrhaphy do not cause de-novo bladder outlet obstruction in the majority of patients with elevated PVR, and the cause of elevated PVR may be other factors such as pain or anxiety causing abnormal relaxation of the pelvic floor and contributing to voiding difficulty. The risk of urinary retention in a future pregnancy after mid-urethral sling (MUS) is small. The risk of urinary tract infection and urinary retention after chemodenervation of the bladder with onabotulinumtoxin-A (100 IU) in patients with non-neurogenic urge incontinence is 33 and 5%, respectively. There is a lack of consensus among experts on the timing of sling takedown in the management of acute urinary retention following MUS procedures. There has been a significant progress in the understanding of the causation of urinary retention. Important areas that need further research (basic and clinical) are post-MUS and pelvic organ prolapse repair urinary retention and obstruction, and urinary retention owing to detrusor underactivity.

In-vessel tritium

International Nuclear Information System (INIS)

Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

2012-01-01

The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

Alignment of in-vessel components by metrology defined adaptive machining

International Nuclear Information System (INIS)

Wilson, David; Bernard, Nathanaël; Mariani, Antony

2015-01-01

Highlights: • Advanced metrology techniques developed for large volume high density in-vessel surveys. • Virtual alignment process employed to optimize the alignment of 440 blanket modules. • Auto-geometry construct, from survey data, using CAD proximity detection and orientation logic. • HMI developed to relocate blanket modules if customization limits on interfaces are exceeded. • Data export format derived for Catia parametric models, defining customization requirements. - Abstract: The assembly of ITER will involve the precise and accurate alignment of a large number of components and assemblies in areas where access will often be severely constrained and where process efficiency will be critical. One such area is the inside of the vacuum vessel where several thousand components shall be custom machined to provide the alignment references for in-vessel systems. The paper gives an overview of the process that will be employed; to survey the interfaces for approximately 3500 components then define and execute the customization process.

Alignment of in-vessel components by metrology defined adaptive machining

Energy Technology Data Exchange (ETDEWEB)

Wilson, David [ITER Organization, Route de Vinon sur Verdon, CS90 046, St Paul-lez-Durance (France); Bernard, Nathanaël [G2Métric, Launaguet 31140 (France); Mariani, Antony [Spatial Alignment Ltd., Witney (United Kingdom)

2015-10-15

Highlights: • Advanced metrology techniques developed for large volume high density in-vessel surveys. • Virtual alignment process employed to optimize the alignment of 440 blanket modules. • Auto-geometry construct, from survey data, using CAD proximity detection and orientation logic. • HMI developed to relocate blanket modules if customization limits on interfaces are exceeded. • Data export format derived for Catia parametric models, defining customization requirements. - Abstract: The assembly of ITER will involve the precise and accurate alignment of a large number of components and assemblies in areas where access will often be severely constrained and where process efficiency will be critical. One such area is the inside of the vacuum vessel where several thousand components shall be custom machined to provide the alignment references for in-vessel systems. The paper gives an overview of the process that will be employed; to survey the interfaces for approximately 3500 components then define and execute the customization process.

Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures

International Nuclear Information System (INIS)

Swindeman, R.W.; Brinkman, C.R.

1981-01-01

Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X

Modification on liquid retention property of cassava starch by radiation grafting with acrylonitrile: Pt. 1

International Nuclear Information System (INIS)

Kiatkamjornwong, S.; Nakason, C.; Chvajarempun, J.

1993-01-01

Radiation modification on liquid retention properties of native cassava starch, gelatinized at 85 o C, by graft copolymerization with acrylonitrile was carried out by mutual irradiation to gamma-rays. A thin aluminium foil was used to cover the inner wall of the reaction vessel, so that the homopolymer concentration was reduced to be less than 1.0% with a distilled water retention value of 665 g/g of the dry weight of the saponified grafted product. Confirmation of graft copolymerization and saponification reactions was made by the infrared spectrophotometric technique. The combined effect of radiation parameters in terms of an irradiation time and a dose rate to the total dose on the extent of the grafting reaction expressed in terms of grafting parameters which directly influenced liquid retention values was evaluated in conjunction with statistical analysis. (author)

Thermal Load Analysis of Multilayered Corium in the Lower Head of Reactor Pressure Vessel during Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Whang, Seok Won; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Hwang, Tae Suk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2014-05-15

In-Vessel Retention (IVR) is one of the severe accident management strategies to terminate or mitigate the severe accident which is also called 'core-melt accident'. The reactor vessel would be cooled by flooding the cavity with water. The molten core mixture is divided into two or three layers due to the density difference. Light metal layer which contains Fe and Zr is on the oxide layer which is consist of UO{sub 2} and ZrO{sub 2}. Heavy metal layer which contains U, Fe and Zr is located under the oxide layer. In oxide layer, the crust which is solidified material is formed along the boundary. The assessment of IVR for nuclear power plant has been conducted with lumped parameter method by Theofanous, Rempe and Esmaili. In this paper, the numerical analysis was performed and verified with the Esmaili's work to analyze thermal load of multilayered corium in pressurized reactor vessel and also to examine the condition of in-vessel corium characteristic before the vessel failure that lead to ex-vessel severe accident progression for example, ex-vessel debris bed cooling. The in-vessel coolability analysis for several scenarios is conducted for the plant which has higher power than AP1000. Two sensitivity analyses are conducted, the first is emissivity of light metal layer and the second is the heat transfer coefficient correlations of oxide layer. The effect of three layered system also investigated. In this paper, the numerical analysis was performed and verified with Esmaili's model to analyze thermal load of multilayered corium in pressurized reactor vessel. For two layered system, thermal load was analyzed according to the severe accident scenarios, emissivity of the light metal layer and heat transfer correlations of the.

Developing retention and return strategies for South African advanced life support paramedics: A qualitative study

Directory of Open Access Journals (Sweden)

Kevin Govender

2013-06-01

Results and discussion: Findings suggest that the success of retention and return strategies depends on the degree of collaborative stewardship that must exist between the SA National and Provincial Department of Health, the Emergency Medical Service/s (EMS statutory governing body, training institutions, and private and public EMS providers. In addition, the constructs of return and retention strategies have to extend as far as revising both acceptance criteria and candidate recruitment policies. Furthermore, while particular attention must be placed on improving working conditions, security, and remuneration of ALS paramedics, return and retention strategies have to be continuously monitored, and updated.

MR angiography of the cerebral vessels with inflow-increased visualization by overcontiguous imaging and advanced processing

International Nuclear Information System (INIS)

McLachian, S.; Simon, H.E.; de Graaf, R.; Ho, A.M.; Ruccio, W.J.; Steidley, J.W.; Pradhan, R.B.

1989-01-01

The purpose of this study was to increase the visualization of cerebral vessels with MR angiography by using an inflow technique from transverse sections. Difficulties with visualization include the tortuous nature of the vessels and slow blood flow. The MR method is a transverse two-dimensional multiple single-section sequence with a velocity-compensated gradient echo and presaturation. A S15 imager with 10-mTm gradients and a prototype angiographic package is used. Overcontiguous acquisition yields smaller effective center-to-center sections. Interpolation is used to generate interleaved plane projections. A combination of image parameter swill reduce the level of the stationary tissue relative to the flowing blood. For vessels that are not perpendicular to the section, overlapping the sections by 25%--33% is shown to reduce the staircase appearance without affecting the apparent size of the vessels. Interpolation of the MR images will further reduce this artifact, giving the vessels a smooth structure. MR angiography of the brain performed with an inflow technique allows the visualization of vessels in the brain even if the vessels follow tortuous paths. Presaturation and user-defined regions of interest can be used for ease of interpretation

Open vessel microwave digestion of food matrices (T6)

International Nuclear Information System (INIS)

Rhodes, L.; LeBlanc, G.

2002-01-01

Full text: Advancements in the field of open vessel microwave digestion continue to provide solutions for industries requiring acid digestion of large sample sizes. Those interesting in digesting food matrices are particularly interested in working with large amounts of sample and then diluting small final volumes. This paper will show the advantages of instantaneous regent addition and post-digestion evaporation when performing an open vessel digestion and evaporation methods for various food matrices will be presented along with analyte recovery data. (author)

Development of the Simulation Program for the In-Vessel Fuel Handling System of Double Rotating Plug Type

International Nuclear Information System (INIS)

Kim, S. H.; Kim, J. B.

2011-01-01

In-vessel fuel handling machines are the main equipment of the in-vessel fuel handling system, which can move the core assembly inside the reactor vessel along with the rotating plug during refueling. The in vessel fuel handling machines for an advanced sodium cooled fast reactor(SFR) demonstration plant are composed of a direct lift machine(DM) and a fixed arm machine(FM). These machines should be able to access all areas above the reactor core by means of the rotating combination of double rotating plugs. Thus, in the in vessel fuel handling system of the double rotating plug type, it is necessary to decide the rotating plug size and evaluate the accessibility of in-vessel fuel handling machines in given core configuration. In this study, the simulation program based on LABVIEW which can effectively perform the arrangement design of the in vessel fuel handling system and simulate the rotating plug motion was developed. Fig. 1 shows the flow chart of the simulation program

Coronary magnetic resonance imaging: visualization of the vessel lumen and the vessel wall and molecular imaging of arteriothrombosis

International Nuclear Information System (INIS)

Spuentrup, Elmar; Botnar, Rene M.

2006-01-01

Coronary magnetic resonance (MR) imaging has dramatically emerged over the last decade. Technical improvements have enabled reliable visualization of the proximal and midportion of the coronary artery tree for exclusion of significant coronary artery disease. However, current technical developments focus also on direct visualization of the diseased coronary vessel wall and imaging of coronary plaque because plaques without stenoses are typically more vulnerable with higher risk of plaque rupture. Plaque rupture with subsequent thrombosis and vessel occlusion is the main cause of myocardial infarction. Very recently, the first success of molecular imaging in the coronary arteries has been demonstrated using a fibrin-specific contrast agent for selective visualization of coronary thrombosis. This demonstrates in general the high potential of molecular MR imaging in the field of coronary artery disease. In this review, we will address recent technical advances in coronary MR imaging, including visualization of the lumen and the vessel wall and molecular imaging of coronary arteriothrombosis. First results of these new approaches will be discussed. (orig.)

Effect of gamma rays on grafting parameters and liquid retention property of cassava starch-g-PAN

International Nuclear Information System (INIS)

Kiatkamjornwong, S.; Chvajarernpun, J.; Nakason, C.

1992-01-01

Radiation modification on liquid retention properties of native cassava starch, gelatinized at 85 0 C, by graft copolymerization with acrylonitrile was carried out by mutual irradiation to γ-rays. A thin aluminium foil was used to cover the inner wall of the reaction vessel so that the extent of homo polymer could be reduced to be less than 1.6% with a distilled water retention value of 665 g/g of the dry weight of the saponified grafted product. Confirmations of graft copolymerization and saponification reactions were made by the infrared spectrophotometric technique. The combined effect of radiation parameters in terms of an irradiation time and a dose rate to the same total dose on the extent of grafting reaction expressed in terms of grafting parameters which directly influenced liquid retention values was evaluated in conjunction with statistical analysis

Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

Science.gov (United States)

Yang, Jun

Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface

ITER vacuum vessel design and electromagnetic analysis on in-vessel components

International Nuclear Information System (INIS)

Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

1995-01-01

Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

Automated method for identification and artery-venous classification of vessel trees in retinal vessel networks.

Science.gov (United States)

Joshi, Vinayak S; Reinhardt, Joseph M; Garvin, Mona K; Abramoff, Michael D

2014-01-01

The separation of the retinal vessel network into distinct arterial and venous vessel trees is of high interest. We propose an automated method for identification and separation of retinal vessel trees in a retinal color image by converting a vessel segmentation image into a vessel segment map and identifying the individual vessel trees by graph search. Orientation, width, and intensity of each vessel segment are utilized to find the optimal graph of vessel segments. The separated vessel trees are labeled as primary vessel or branches. We utilize the separated vessel trees for arterial-venous (AV) classification, based on the color properties of the vessels in each tree graph. We applied our approach to a dataset of 50 fundus images from 50 subjects. The proposed method resulted in an accuracy of 91.44% correctly classified vessel pixels as either artery or vein. The accuracy of correctly classified major vessel segments was 96.42%.

Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

International Nuclear Information System (INIS)

Foerthmann, R.; Groos, E.; Gruebmeier, H.

1975-08-01

Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention of the kernel, which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl 2 Si 2 O 8 , BaAl 2 Si 2 O 8 and CsAlSi 2 O 6 in the additional aluminasilica phase of the kernel. (orig.) [de

Enhancing retention of partial dentures using elastomeric retention rings

Directory of Open Access Journals (Sweden)

Kakkirala Revathi

2015-01-01

Full Text Available This report presents an alternative method for the retention of partial dentures that relies on the engagement of tooth undercuts by a lining material. The lab procedures are also presented. A new maxillary and mandibular acrylic partial dentures were fabricated using elastomeric retention technique for a partially dentate patient. A partially dentate man reported difficulty in retaining his upper removable partial denture (RPD. The maxillary RPD was designed utilizing elastomeric retention technique. During follow-up, it was necessary to replace the retention rings due to wear. The replacement of the retention rings, in this case, was done through a chairside reline technique. Elastomeric retention technique provides exceptionally good retention can be indicated to stabilize, cushion, splint periodontally involved teeth, no enough undercut for clasps, eliminate extractions, single or isolated teeth.

Lace-Espana experimental programme on the retention of aerosols in water pools

Energy Technology Data Exchange (ETDEWEB)

Marcos, M. J.; Gomez, F. J.; Melches, I.; Martin, M.; Lopez, M.

1994-07-01

A matrix of eleven experiments on aerosol retention behaviour in submerged beds and suppression pools in water- cooled reactors under severe accident conditions has been performed, for these experiments, an intermediate scales, multi-purpose facility was set up at CIEMAT (Madrid). The facility includes various systems: aerosol generation (Csl), mixing section, injection line and pool-vessel (8 m{sup 3} ), as well as the corresponding aerosol instrumentation and a process control and data acquisition system. Some parameters have been varied in order to study their influence in the DF: steam/noncondensable ratio in the accidental mixture (0.1 to 0.9), particle size, flow rate (two regimes: bubble and jet) and injector geometry (mono orifice and multi orifice). On the other hand, some parameters have been kept constant along the experiments; pool geometry (diameter, water level), water temperature, pressure in the atmosphere above the water, submergence, injection temperature and injection time. A rapid decrease in the DF is observed as the proportion of particles measuring less than 1 {mu}m increases. Retention decreases in the case of smaller particles and considerably higher in the case of larger particles. It has been also possible to observe the influence of the injected steam fraction. Experiments with greater fraction than the saturation fraction have greater DF than those ones with smaller fractions. The jet regime with horizontal injection and the multi orifice geometry would appear to show a somewhat higher capacity of retention than those in the bubble regime under similar conditions. It would be necessary to confirm this greater capacity for retention by means of additional experimental data. This work, performed by the LACE-Espana Consortium, has been carried out in the frame of the European Commissions Shared Cost Action Programme on Reactor Safety 1988-91 on a contractual basis. (Author)18 refs.

Lace-Espana experimental programme on the retention of aerosols in water pools

International Nuclear Information System (INIS)

Marcos, M. J.; Gomez, F. J.; Melches, I.; Martin, M.; Lopez, M.

1994-01-01

A matrix of eleven experiments on aerosol retention behaviour in submerged beds and suppression pools in water- cooled reactors under severe accident conditions has been performed, for these experiments, an intermediate scales, multi-purpose facility was set up at CIEMAT (Madrid). The facility includes various systems: aerosol generation (Csl), mixing section, injection line and pool-vessel (8 m 3 ), as well as the corresponding aerosol instrumentation and a process control and data acquisition system. Some parameters have been varied in order to study their influence in the DF: steam/noncondensable ratio in the accidental mixture (0.1 to 0.9), particle size, flow rate (two regimes: bubble and jet) and injector geometry (mono orifice and multi orifice). On the other hand, some parameters have been kept constant along the experiments; pool geometry (diameter, water level), water temperature, pressure in the atmosphere above the water, submergence, injection temperature and injection time. A rapid decrease in the DF is observed as the proportion of particles measuring less than 1 μm increases. Retention decreases in the case of smaller particles and considerably higher in the case of larger particles. It has been also possible to observe the influence of the injected steam fraction. Experiments with greater fraction than the saturation fraction have greater DF than those ones with smaller fractions. The jet regime with horizontal injection and the multi orifice geometry would appear to show a somewhat higher capacity of retention than those in the bubble regime under similar conditions. It would be necessary to confirm this greater capacity for retention by means of additional experimental data. This work, performed by the LACE-Espana Consortium, has been carried out in the frame of the European Commissions Shared Cost Action Programme on Reactor Safety 1988-91 on a contractual basis. (Author)18 refs

Lace-Espana experimental programme on the retention of aerosols in water pools

International Nuclear Information System (INIS)

Marcos Crespo, M.J.; Gomez; Moreno, F. J.; Melches Serrano, I.; Martin Espigares, M.; Lopez Jimenez, J.

1994-01-01

A matrix of eleven experiments on aerosol retention behaviour in submerged beds and suppression pools in water-cooled reactors under severe accident conditions has been performed, for these experiments, an intermediate scale, multi-purpose facility was set up at CIEMAT (Madrid). The facility includes various systems: aerosol ageneration (CsI), mixing section, injection line and pool-vessel (8 m''3), as well as the corresponding aerosol instrumentation and a process control and data acquisition system. Some parameters have been varied in order to study their influence in the DF: steam/noncondensable ratio in the accidental mixture (0,1 to 0.9) particle size, flow rate (two regimes: bubble and jet) and injector geometry (monoorifice and multiorifice). On the other hand, some parameters have been kept constant along the experiments; pool geometry (diameter, water level), water temperature, pressure in the atmosphere above the water, submergence, injection temperature and injection time. A Rapid decrease in the DF is observed as the proportion of particles measuring les than 1 mum increases. Retention decreases in the case of smaller particles and considerably higher in the case of larger particles. It has been also possible to observe the influence of the injected steam fraction. Experiments with greater fraction than the saturation fraction have greater DF than those ones with smaller fractions. The jet regime with horizontal injection and the multiorifice geometry would appear to show a somewhat higher capacity of retention than those in the bubble regime under similar conditions. It would be necessary to confirm this greater capacity for retention by means of additional experimental data. This work, performed by the LACE-Espana Consortium, has been carried out in the frame of the European Commission's Shared Cost Action Programme on Reactor Safety 1988-91 on a contractual basis

Employee retention: a customer service approach.

Science.gov (United States)

Gerson, Richard F

2002-01-01

Employee retention is a huge problem. There are staff shortages in radiology because not enough people are entering the profession; too many people are leaving the profession for retirement, higher-paying jobs or jobs with less stress; and there are not enough opportunities for career advancement. Staff shortages are exacerbated by difficulty in retaining people who enter the profession. While much work has been focused on recruitment and getting more people "in the front door," I suggest that the bulk of future efforts be focused on employee retention and "closing the back door." Employee retention must be an ongoing process, not a program. Approaches to employee retention that focus on external things, i.e., things that the company can do to or for the employee, generally are not successful. The truth is that employee retention processes must focus on what the employee gets out of the job. The process must be a benefits-based approach that helps employees answer the question, "What's in it for me?" The retention processes must be ongoing and integrated into the daily culture of the company. The best way to keep your employees is to treat them like customers. Customer service works for external customers. We treat them nicely. We work to satisfy them. We help them achieve their goals. Why not do the same for our employees? If positive customer service policies and practices can satisfy and keep external customers, why not adapt these policies and practices for employees? And, there is a service/satisfaction link between employee retention and higher levels of customer satisfaction. Customers prefer dealing with the same employees over and over again. Employee turnover destroys a customer's confidence in the company. Just like a customer does not want to have to "train and educate" a new provider, they do not want to do the same for your "revolving door" employees. So, the key is to keep employees so they in turn will help you keep your customers. Because the

An analysis of critical heat flux on the external surface of the reactor vessel lower head

International Nuclear Information System (INIS)

Yang, Soo Hyung; Baek, Won Pil; Chang, Soon Heung

1999-01-01

CHF (Critical heat flux) on the external surface of the reactor vessel lower head is major key in the evaluation on the feasibility of IVR-EVC (In-Vessel Retention through External Vessel Cooling) concept. To identify the CHF on the external surface, considerable works have been performed. Through the review on the previous works related to the CHF on the external surface, liquid subcooling, induced flow along the external surface, ICI (In-Core Instrument) nozzle and minimum gap are identified as major parameters. According to the present analysis, the effects of the ICI nozzle and minimum gap on CHF are pronounced at the upstream of test vessel: on the other hand, the induced flow considerably affects the CHF at downstream of test vessel. In addition, the subcooling effect is shown at all of test vessel, and decreases with the increase in the elevation of test vessel. In the real application of the IVR-EVC concept, vertical position is known as a limiting position, at which thermal margin is the minimum. So, it is very important to precisely predict the CHF at vertical position in a viewpoint of gaining more thermal margins. However, the effects of the liquid subcooling and induced flow do not seem to be adequately included in the CHF correlations suggested by previous works, especially at the downstream positions

Cordless ultrasonic dissector versus advanced bipolar vessel sealing device for laparoscopic ovariectomy in dogs

NARCIS (Netherlands)

Spillebeen, Anneleen L; Janssens, Sara S D S; Thomas, Rachel E; Kirpensteijn, Jolle; van Nimwegen, Sebastiaan A

OBJECTIVE: To compare Sonicision cordless ultrasonic dissector (SCUD) to LigaSure vessel sealing device (LVSD) for laparoscopic ovariectomy (Lap OVE) in dogs. STUDY DESIGN: Randomized, paired prospective clinical trial. ANIMALS: Client-owned dogs (n = 22) presented for elective Lap OVE. METHODS:

Arctic research vessel design would expand science prospects

Science.gov (United States)

Elsner, Robert; Kristensen, Dirk

The U.S. polar marine science community has long declared the need for an arctic research vessel dedicated to advancing the study of northern ice-dominated seas. Planning for such a vessel began 2 decades ago, but competition for funding has prevented construction. A new design program is underway, and it shows promise of opening up exciting possibilities for new research initiatives in arctic marine science.With its latest design, the Arctic Research Vessel (ARV) has grown to a size and capability that will make it the first U.S. academic research vessel able to provide access to the Arctic Ocean. This ship would open a vast arena for new studies in the least known of the world's seas. These studies promise to rank high in national priority because of the importance of the Arctic Ocean as a source of data relating to global climate change. Other issues that demand attention in the Arctic include its contributions to the world's heat budget, the climate history buried in its sediments, pollution monitoring, and the influence of arctic conditions on marine renewable resources.

An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

Energy Technology Data Exchange (ETDEWEB)

Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

2004-01-01

LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

Fuel hydrogen retention of tungsten and the reduction by inert gas glow discharges

Energy Technology Data Exchange (ETDEWEB)

Hino, T., E-mail: tomhino@qe.eng.hokudai.ac.jp [Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Yamauchi, Y.; Kimura, Y. [Laboratory of Plasma Physics and Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Nishimura, K. [National Institute for Fusion Science, Toki-shi, Gifu-ken 509-5292 (Japan); Ueda, Y. [Graduate School of Engineering, Osaka University, Suita-shi 565-0872 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The performances of inert gas glow discharges for reduction of fuel hydrogen retention in tungsten were systematically investigated. Black-Right-Pointing-Pointer For the tungsten with rough surface structure, the reduction of fuel hydrogen retention by inert gas discharges is quite small. Black-Right-Pointing-Pointer The deuterium glow discharge is quite useful to reduce the tritium retention in plasma facing walls in fusion reactor. Black-Right-Pointing-Pointer The wall baking with temperature higher than 700-800 K is also useful to reduce the tritium retention in plasma facing walls. - Abstract: Polycrystalline tungsten was exposed to deuterium glow discharge followed by He, Ne or Ar glow discharge. The amount of retained deuterium in the tungsten was measured using residual gas analysis. The amount of desorbed deuterium during the inert gas glow discharge was also measured. The amount of retained deuterium was 2-3 times larger compared with a case of stainless steel. The ratios of desorbed amount of deuterium by He, Ne and Ar glow discharges were 4.6, 3.1 and 2.9%, respectively. These values were one order of magnitude smaller compared with the case of stainless steel. The inert gas glow discharge is not suitable to reduce the fuel hydrogen retention for tungsten walls. However, the wall baking with a temperature higher than 700 K is suitable to reduce the fuel hydrogen retention. It is also shown that the use of deuterium glow discharge is effective to reduce the in-vessel tritium inventory in fusion reactors through the hydrogen isotope exchange.

Advances in radiological imaging of the renal arteries and veins

International Nuclear Information System (INIS)

Miao Xiaofan; Tang Lijun; Yang Bing

2013-01-01

Familiarity with the normal anatomy of the renal vessels and common variants is of particular importance for the operator who performs renal transplantation or therapeutic interventions in the renal vessels. Because of the recent major advances in multislice spiral computed tomography angiography (MSCTA) techniques, our ability has been considerably improved to determine the patterns and characters of renal arteries and veins. This article summarizes the research situation and progress in the area of renal vessel imaging anatomy. Some regularity in the distribution of renal vessel positions, dimensions and variations among patients who received MSCTA examinations, and these anatomical measurements are of great value for clinical diagnosis and treatment. (authors)

In-calandria retention of corium in Indian PHWR - experimental simulations with decay heat

International Nuclear Information System (INIS)

Nayak, A.K.

2015-01-01

The severe accident at Fukushima has compelled the nuclear community to relook at the safety of existing nuclear power plants (NPP) against natural origin events of beyond design basis and prolonged station black out (SBO). A major lesson learned is to assess the capability of the safety systems to cool the reactor core and spent fuel storage facilities in the event of a prolonged station black out (SBO). Similar safety review is planned for the Indian Pressurized Heavy Water Reactors (PHWRs) considering a prolonged SBO. The Indian PHWR is a heavy water-moderated and cooled, natural uranium-fuelled reactor in which the horizontal fuel channels are submerged in a pool of heavy water moderator located inside the calandria vessel. The calandria vessel is surrounded by a calandria vault having large volume of light water. Concerns are raised that in the event of an unmitigated SBO, it may result into a low probable severe accident leading to core melt down. The core melt may further fail the calandria vessel in case the melt is not quenched. If the calandria vessel fails, the corium shall interact with the cold calandria vault water and concrete resulting in generation of large amount of non-condensable gases and steam which will lead to over pressurization of containment and may cause its failure. Therefore, in-calandria corium retention via external cooling using vault water can be considered as an important accident management program in PHWR. In this strategy, the core melt retains inside the calandria vessel by continually removing the stored heat and decay heat through outer surface of the vessel by cooling water and maintaining the integrity of the vessel. The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat by using the calandria vault water. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics

Survival Analysis of Faculty Retention and Promotion in the Social Sciences by Gender.

Directory of Open Access Journals (Sweden)

Janet M Box-Steffensmeier

Full Text Available Recruitment and retention of talent is central to the research performance of universities. Existing research shows that, while men are more likely than women to be promoted at the different stages of the academic career, no such difference is found when it comes to faculty retention rates. Current research on faculty retention, however, focuses on careers in science, technology, engineering, and mathematics (STEM. We extend this line of inquiry to the social sciences.We follow 2,218 tenure-track assistant professors hired since 1990 in seven social science disciplines at nineteen U.S. universities from time of hire to time of departure. We also track their time to promotion to associate and full professor. Using survival analysis, we examine gender differences in time to departure and time to promotion. Our methods account for censoring and unobserved heterogeneity, as well as effect heterogeneity across disciplines and cohorts.We find no statistically significant differences between genders in faculty retention. However, we do find that men are more likely to be granted tenure than women. When it comes to promotion to full professor, the results are less conclusive, as the effect of gender is sensitive to model specification.The results corroborate previous findings about gender patterns in faculty retention and promotion. They suggest that advances have been made when it comes to gender equality in retention and promotion, but important differences still persist.

Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

International Nuclear Information System (INIS)

Onozuka, Masanori; Nakahira, Masataka

2006-01-01

The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

Coarse Particulate Organic Matter: Storage, Transport, and Retention

Energy Technology Data Exchange (ETDEWEB)

Tiegs, Scott [Oakland University, Rochester, MI; Lamberti, Gary A. [University of Notre Dame, IN; Entrekin, Sally A. [University of Central Arkansas; Griffiths, Natalie A. [ORNL

2017-08-01

Coarse particulate organic matter, or CPOM, is a basal energy and nutrient resource in many stream ecosystems and is provided by inputs from the riparian zone, incoming tributaries, and to a lesser extent from in-stream production. The ability of a stream to retain CPOM or slow its transport is critical to its consumption and assimilation by stream biota. In this chapter, we describe basic exercises to measure (1) the amount of CPOM in the streambed and (2) the retention of CPOM from standardized particle releases. We further describe advanced exercises that (1) experimentally enhance the retentiveness of a stream reach and (2) measure organic carbon transport and turnover (i.e., spiraling) in the channel.

Coarse Particulate Organic Matter: Storage, Transport, and Retention

Energy Technology Data Exchange (ETDEWEB)

Tiegs, Scott [Oakland University, Rochester, MI; Lamberti, Gary A. [University of Notre Dame, IN; Entrekin, Sally A. [University of Central Arkansas; Griffiths, Natalie A. [ORNL

2017-06-01

Coarse particulate organic matter, or CPOM, is a basal energy and nutrient resource in many stream ecosystems and is provided by inputs from the riparian zone, incoming tributaries, and to a lesser extent from in-stream production. The ability of a stream to retain CPOM or slow its transport is critical to its consumption and assimilation by stream biota. In this chapter, we describe basic exercises to measure (1) the amount of CPOM in the streambed and (2) the retention of CPOM from standardized particle releases. We further describe advanced exercises that (1) experimentally enhance the retentiveness of a stream reach and (2) measure organic carbon transport and turnover (i.e., spiraling) in the channel.

Tumor Vessel Compression Hinders Perfusion of Ultrasonographic Contrast Agents

Directory of Open Access Journals (Sweden)

Mirco Galiè

2005-05-01

Full Text Available Contrast-enhanced ultrasound (CEUS is an advanced approach to in vivo assessment of tumor vascularity and is being increasingly adopted in clinical oncology. It is based on 1- to 10 μm-sized gas microbubbles, which can cross the capillary beds of the lungs and are effective echo enhancers. It is known that high cell density, high transendothelial fluid exchange, and poorly functioning lymphatic circulation all provoke solid stress, which compresses vessels and drastically reduces tumor blood flow. Given their size, we supposed that the perfusion of microbubbles is affected by anatomic features of tumor vessels more than are contrast agents traditionally used in dynamic contrast-enhanced magnetic resonance imaging (DCE-MRI. Here, we compared dynamic information obtained from CEUS and DCE-MRI on two experimental tumor models exhibiting notable differences in vessel anatomy. We found that tumors with small, flattened vessels show a much higher resistance to microbubble perfusion than to MRI contrast agents, and appear scarcely vascularized at CEUS examination, despite vessel volume adequate for normal function. Thus, whereas CEUS alone could induce incorrect diagnosis when tumors have small or collapsed vessels, integrated analysis using CEUS and DCE-MRI allows in vivo identification of tumors with a vascular profile frequently associated with malignant phenotypes.

Improvement in retention of solid fission products in HTGR fuel particles by ceramic kernel additives

Energy Technology Data Exchange (ETDEWEB)

Foerthmann, R.; Groos, E.; Gruebmeier, H.

1975-08-15

Increased requirements concerning the retention of long-lived solid fission products in fuel elements for use in advanced High Temperature Gas-cooled Reactors led to the development of coated particles with improved fission product retention which represent an alternative to silicon carbide-coated fuel particles. Two irradiation experiments have shown that the release of strontium, barium, and caesium from pyrocarbon-coated particles can be reduced by orders of magnitude if the oxide kernel contains alumina-silica additives. It was detected by electron microprobe analysis that the improved retention of the mentioned fission products in the fuel kernel is caused by formation of the stable aluminosilicates SrAl₂Si₂O₈, BaAl₂Si₂O₈and CsAlSi₂O₆ in the additional alumina-silica phase of the kernel.

Design Procedure on Stud Bolt for Reactor Vessel Assembly

International Nuclear Information System (INIS)

Kim, Jong-Wook; Lee, Gyu-Mahn; Jeoung, Kyeong-Hoon; Kim, Tae-Wan; Park, Keun-Bae; Kim, Keung-Koo

2008-10-01

The reactor pressure vessel flange is welded to the upper part of reactor pressure vessel, and there are stud holes to mount the closure head with stud bolts. The surface mating the closure head is compressed with O-ring, which acts as a sealing gasket to prevent coolant leakage. Bolted flange connections perform a very important structural role in the design of a reactor pressure vessel. Their importance stems from two important functions: (a) maintenance of the structural integrity of the connection itself, and (b) prevention of leakage through the O-ring preloaded by stud bolts. In the present study, an evaluation procedure for the design of stud bolt is developed to meet ASME code requirements. The developed design procedure could provide typical references in the development of advanced reactor design in the future

FOREVER Experiments on Thermal and Mechanical Behavior of a Reactor Pressure Vessel during a Severe Accident

International Nuclear Information System (INIS)

Sehgal, B.R.; Nourgaliev, R.R.; Dinh, T.N.; Karbojian, A.; Green, J.A.; Bui, V.A.

1999-01-01

This paper describes the FOREVER (Failure Of Reactor Vessel Retention) experimental program, which is currently underway at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS). The objectives of the FOREVER experiments are to obtain data and develop validated models (i) on the melt coolability process inside the vessel, in the presence of water (in particular, on the efficacy of the postulated gap cooling to preclude vessel failure); and (ii) on the lower head failure due to the creep process in the absence of water inside and/or outside the lower head. The facility employs 1/10.-scale carbon steel vessels of 0.4 m diameter, 15 mm thickness and 600 mm height. Up to 20 liters of binary-oxide melts with 100-300 K superheat are employed, as a simulant for the prototypic corium melt, and internal heating is provided by electrical heaters of up to 20 kW power in order to maintain the vessel wall temperatures at 1100-1200 K. Auxiliary systems are designed to provide an overpressure up to 4 MPa in the test vessel. Thus, severe accident scenarios with RCS depressurization are modeled. Creep behavior of the three-dimensional vessel, formation of the gap between the melt pool crust and the creeping vessel, and mechanisms of the gap cooling by water ingression will be the subjects of study and measurements in the FOREVER experimental program. Scaling rationale as well as pre-test analyses of the thermal and mechanical behavior of the FOREVER test vessels are presented. (authors)

Vessel classification method based on vessel behavior in the port of Rotterdam

NARCIS (Netherlands)

Zhou, Y.; Daamen, W.; Vellinga, T.; Hoogendoorn, S.P.

2015-01-01

AIS (Automatic Identification System) data have proven to be a valuable source to investigate vessel behavior. The analysis of AIS data provides a possibility to recognize vessel behavior patterns in a waterway area. Furthermore, AIS data can be used to classify vessel behavior into several

Requirements for thermal insulation on prestressed concrete reactor vessels

International Nuclear Information System (INIS)

Neylan, A.J.; Wistrom, J.D.

1979-01-01

During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

The Development of Key Technologies in Applications of Vessels Connected to the Internet

Directory of Open Access Journals (Sweden)

Zhe Tian

2017-10-01

Full Text Available With the development of science and technology, traffic perception, communication, information processing, artificial intelligence and the shipping information system have become important in supporting the realization of intelligent shipping transportation. Against this background, the Internet of Vessels (IoV is proposed to integrate all these advanced technologies into a platform to meet the requirements of international and regional transportations. The purpose of this paper is to analyze how to benefit from the Internet of Vessels to improve the efficiency and safety of shipping, and promote the development of world transportation. In this paper, the IoV is introduced and its main architectures are outlined. Furthermore, the characteristics of the Internet of Vessels are described. Several important applications that illustrate the interaction of the Internet of Vessels’ components are proposed. Due to the development of the Internet of Vessels still being in its primary stage, challenges and prospects are identified and addressed. Finally, the main conclusions are drawn and future research priorities are provided for reference and as professional suggestions for future researchers in this field.

INETEC new system for inspection of PWR reactor pressure vessel head

International Nuclear Information System (INIS)

Nadinic, B.; Postruzin, Z.

2004-01-01

INETEC Institute for Nuclear Technology developed new equipment for inspection of PWR and VVER reactor pressure vessel head. The new advances in inspection technology are presented in this article, as the following: New advance manipulator for inspection of RPVH with high speed of inspection possibilities and total automated work; New sophisticated software for manipulator driving which includes 3D virtual presentation of manipulator movement and collision detection possibilities; New multi axis controller MAC-8; New end effector system for inspection of penetration tube and G weld; New eddy current and ultrasonic probes for inspection of G weld and penetration tube; New Eddy One Raster scan software for analysis of eddy current data with mant advanced features which allows easy and quick data analysis. Also the results of laboratory testing and laboratory qualification are presented on reactor pressure vessel head mock, as well as obtained speed of inspection and quality of collected data.(author)

The TPX vacuum vessel and in-vessel components

International Nuclear Information System (INIS)

Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

1994-01-01

The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

TECHcitement: Advances in Technology Education, 2008

Science.gov (United States)

Patton, Madeline

2008-01-01

This publication presents the following articles: (1) Advanced Technological Education (ATE) Develops Student Recruitment and Retention Strategies; (2) Marketer Advises Tech Educators Appeal to Teens' Emotions, Desires to Do Something Important; (3) Digital Bridge Academy Gets At-Risk Students on Paths to Knowledge-Based Careers; (4) Project…

Urinary retention in women

African Journals Online (AJOL)

Urinary retention in women. Urinary retention in women is often transient and of no known cause. ... stones, constipation, urethral cancer, uterine fibroids ... present with abnormal bladder function secondary to ... (CT), magnetic resonance imaging (MRI) or myelography ... full blood count, urea, electrolytes and creatinine can ...

Blended Retirement System Impact on Gender Retention

Science.gov (United States)

2017-04-13

compensation in their decisions on how long to remain in the service, but financial compensation is not the most important factor. In order to continue...their Thrift Savings Plan (TSP) retirement account to provide members departing the service prior to 20 year retirement eligibility with some portable...employment and promotion/ advancement opportunities. Thus, the BRS will have a neutral effect on gender 4 retention. In order to continue to close

Primo vessel inside a lymph vessel emerging from a cancer tissue.

Science.gov (United States)

Lee, Sungwoo; Ryu, Yeonhee; Cha, Jinmyung; Lee, Jin-Kyu; Soh, Kwang-Sup; Kim, Sungchul; Lim, Jaekwan

2012-10-01

Primo vessels were observed inside the lymph vessels near the caudal vena cava of a rabbit and a rat and in the thoracic lymph duct of a mouse. In the current work we found a primo vessel inside the lymph vessel that came out from the tumor tissue of a mouse. A cancer model of a nude mouse was made with human lung cancer cell line NCI-H460. We injected fluorescent nanoparticles into the xenografted tumor tissue and studied their flow in blood, lymph, and primo vessels. Fluorescent nanoparticles flowed through the blood vessels quickly in few minutes, and but slowly in the lymph vessels. The bright fluorescent signals of nanoparticles disappeared within one hour in the blood vessels but remained much longer up to several hours in the case of lymph vessels. We found an exceptional case of lymph vessels that remained bright with fluorescence up to 24 hours. After detailed examination we found that the bright fluorescence was due to a putative primo vessel inside the lymph vessel. This rare observation is consistent with Bong-Han Kim's claim on the presence of a primo vascular system in lymph vessels. It provides a significant suggestion on the cancer metastasis through primo vessels and lymph vessels. Copyright © 2012. Published by Elsevier B.V.

External cooling: The SWR 1000 severe accident management strategy. Part 1: motivation, strategy, analysis: melt phase, vessel integrity during melt-water interaction

International Nuclear Information System (INIS)

Kolev, Nikolay Ivanov

2004-01-01

This paper provides the description of the basics behind design features for the severe accident management strategy of the SWR 1000. The hydrogen detonation/deflagration problem is avoided by containment inertization. In-vessel retention of molten core debris via water cooling of the external surface of the reactor vessel is the severe accident management concept of the SWR 1000 passive plant. During postulated bounding severe accidents, the accident management strategy is to flood the reactor cavity with Core Flooding Pool water and to submerge the reactor vessel, thus preventing vessel failure in the SWR 1000. Considerable safety margins have determined by using state of the art experiment and analysis: regarding (a) strength of the vessel during the melt relocation and its interaction with water; (b) the heat flux at the external vessel wall; (c) the structural resistance of the hot structures during the long term period. Ex-vessel events are prevented by preserving the integrity of the vessel and its penetrations and by assuring positive external pressure at the predominant part of the external vessel in the region of the molten corium pool. Part 1 describes the motivation for selecting this strategy, the general description of the strategy and the part of the analysis associated with the vessel integrity during the melt-water interaction. (author)

Numerical simulation of multi-phase phenomena in IVR related processes

Energy Technology Data Exchange (ETDEWEB)

Cheng, Xu [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany). Bereich Innovative Reaktorsysteme; Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering

2016-05-15

IVR (in-vessel retention) is one of the severe accident mitigation measures, which is widely applied in the advanced light water reactors (LWRs) such as KERENA of AREVA, AP1000 of Westinghouse and CAP1400 of SNPTC, and attracts extensive interests of the German and Chinese nuclear scientists. The ultimate target of IVR is to keep the core melt inside the reactor pressure vessel (RPV) and to provide cooling capability via water flowing outside the RPV, the so called external reactor vessel cooling (ERVC). This paper summarizes some activities ongoing in both KIT and SJTU (Shanghai Jiao Tong University) with a few results examples.

Recent advances in vibro-impact dynamics and collision of ocean vessels

Science.gov (United States)

Ibrahim, Raouf A.

2014-11-01

The treatment of ship impacts and collisions takes different approaches depending on the emphasis of each discipline. For example, dynamicists, physicist, and mathematicians are dealing with developing analytical models and mappings of vibro-impact systems. On the other hand, naval architects and ship designers are interested in developing design codes and structural assessments due to slamming loads, liquid sloshing impact loads in liquefied natural gas tanks and ship grounding accidents. The purpose of this review is to highlight the main differences of the two disciplines. It begins with a brief account of the theory of vibro-impact dynamics based on modeling and mapping of systems experiencing discontinuous changes in their state of motion due to collision. The main techniques used in modeling include power-law phenomenological modeling, Hertzian modeling, and non-smooth coordinate transformations originally developed by Zhuravlev and Ivanov. In view of their effectiveness, both Zhuravlev and Ivanov non-smooth coordinate transformations will be described and assessed for the case of ship roll dynamics experiencing impact with rigid barriers. These transformations have the advantage of converting the vibro-impact oscillator into an oscillator without barriers such that the corresponding equation of motion does not contain any impact term. One of the recent results dealing with the coefficient of restitution is that its value monotonically decreases with the impact velocity and not unique but random in nature. Slamming loads and grounding events of ocean waves acting on the bottom of high speed vessels will be assessed with reference to the ship structural damage. It will be noticed that naval architects and marine engineers are treating these problems using different approaches from those used by dynamicists. The problem of sloshing impact in liquefied natural gas cargo and related problems will be assessed based on the numerical and experimental results. It is

Engineering blood vessels through micropatterned co-culture of vascular endothelial and smooth muscle cells on bilayered electrospun fibrous mats with pDNA inoculation.

Science.gov (United States)

Liu, Yaowen; Lu, Jinfu; Li, Huinan; Wei, Jiaojun; Li, Xiaohong

2015-01-01

Although engineered blood vessels have seen important advances during recent years, proper mechanical strength and vasoactivity remain unsolved problems. In the current study, micropatterned fibrous mats were created to load smooth muscle cells (SMC), and a co-culture with endothelial cells (EC) was established through overlaying on an EC-loaded flat fibrous mat to mimic the layered structure of a blood vessel. A preferential distribution of SMC was determined in the patterned regions throughout the fibrous scaffolds, and aligned fibers in the patterned regions provided topological cues to guide the orientation of SMC with intense actin filaments and extracellular matrix (ECM) production in a circumferential direction. Plasmid DNA encoding basic fibroblast growth factors and vascular endothelial growth factor were integrated into electrospun fibers as biological cues to promote SMC infiltration into fibrous mats, and the viability and ECM production of both EC and SMC. The layered fibrous mats with loaded EC and SMC were wrapped into a cylinder, and engineered vessels were obtained with compact EC and SMC layers after co-culture for 3 months. Randomly oriented ECM productions of EC formed a continuous endothelium covering the entire lumenal surface, and a high alignment of ECM was shown in the circumferential direction of SMC layers. The tensile strength, strain at failure and suture retention strength were higher than those of the human femoral artery, and the burst pressure and radial compliance were in the same range as the human saphenous vein, indicating potential as blood vessel substitutes for transplantation in vivo. Thus, the establishment of topographical cues and biochemical signals in fibrous scaffolds demonstrates advantages in modulating cellular behavior and organization found in complex multicellular tissues. Copyright © 2014 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

Mass transfer experiments for the heat load during in-vessel retention of core melt

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Chung, Bum Jin [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

2016-08-15

We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a CuSO{sub 4}–H{sub 2}SO{sub 4} electroplating system were performed based on the analogy between heat and mass transfer. The Ra′{sub H} of 10{sup 14} order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.

Interaction between molten corium UO2+x-ZrO2-FeOy and VVER vessel steel

International Nuclear Information System (INIS)

Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A.; Gusarov, V. V.; Almiashev, V. I.; Lopukh, D. B.; Bottomley, D.; Fischer, M.; Piluso, P.; Miassoedov, A.; Tromm, W.; Altstadt, E.; Fichot, F.; Kymalainen, O.

2010-01-01

In case of in-vessel corium retention during a severe accident in a light water reactor, weakening of the vessel wall and deterioration of the vessel steel properties can be caused both by the melting of the steel and by its physicochemical interaction with corium. The interaction behavior has been studied in medium-scale experiments with prototypic corium. The experiments yielded data for the steel corrosion rate during interaction with UO 2+x -ZrO 2 -FeO y melt in air and steam at different steel surface temperatures and heat fluxes from the corium to the steel. It has been observed that the corrosion rates in air and steam atmosphere are almost the same. Further, if the temperature at the interface increases beyond a certain level, corrosion intensifies. This is explained by the formation of liquid phases in the interaction Zone. The available experimental data have been used to develop a correlation for the corrosion rate as a function of temperature and heat flux. (authors)

Tumor Vessel Compression Hinders Perfusion of Ultrasonographic Contrast Agents1

Science.gov (United States)

Galiè, Mirco; D'Onofrio, Mirko; Montani, Maura; Amici, Augusto; Calderan, Laura; Marzola, Pasquina; Benati, Donatella; Merigo, Flavia; Marchini, Cristina; Sbarbati, Andrea

2005-01-01

Abstract Contrast-enhanced ultrasound (CEUS) is an advanced approach to in vivo assessment of tumor vascularity and is being increasingly adopted in clinical oncology. It is based on 1- to 10 µm-sized gas microbubbles, which can cross the capillary beds of the lungs and are effective echo enhancers. It is known that high cell density, high transendothelial fluid exchange, and poorly functioning lymphatic circulation all provoke solid stress, which compresses vessels and drastically reduces tumor blood flow. Given their size, we supposed that the perfusion of microbubbles is affected by anatomic features of tumor vessels more than are contrast agents traditionally used in dynamic contrast-enhanced magnetic resonance imaging (DCE-MRI). Here, we compared dynamic information obtained from CEUS and DCE-MRI on two experimental tumor models exhibiting notable differences in vessel anatomy. We found that tumors with small, flattened vessels show a much higher resistance to microbubble perfusion than to MRI contrast agents, and appear scarcely vascularized at CEUS examination, despite vessel volume adequate for normal function. Thus, whereas CEUS alone could induce incorrect diagnosis when tumors have small or collapsed vessels, integrated analysis using CEUS and DCE-MRI allows in vivo identification of tumors with a vascular profile frequently associated with malignant phenotypes. PMID:15967105

Polylysine as a vehicle for extracellular matrix-targeted local drug delivery, providing high accumulation and long-term retention within the vascular wall

NARCIS (Netherlands)

Sakharov, D.V.; Jie, A.F.H.; Bekkers, M.E.A.; Emeis, J.J.; Rijken, D.C.

2001-01-01

We present the first steps in the elaboration of an approach of extracellular matrix-targeted local drug delivery (ECM-LDD), designed to provide a high concentration, ubiquitous distribution, and long-term retention of a drug within the vessel wall after local intravascular delivery. The approach is

A method of decontaminating large vessels

International Nuclear Information System (INIS)

Williams, John.

1989-01-01

A method of removing contamination from the surfaces of nuclear reactors, heat exchangers and pressure vessels which requires the use of only small volumes of liquid and no abrasives involves contacting the contaminated surface with a moisture laden gas and effecting condensation of the moisture in the gas on the surface by lowering the temperature of the surface to be cleaned before introducing the moisture laden gas. A reactive cleaning agent is concurrently or subsequently introduced and allowed to react with the contaminated surface to produce a water soluble reaction product which is washed away. To ensure the retention of a wetted surface for reactant contact and to avoid loss of reactant contact by evaporation or drainage at the surface a modification of the method provides for the production of the gelatinous layer on the surface to be cleaned. (author)

Draft paper: On the analysis of diffusive mass transfer in ex-vessel corium pools

International Nuclear Information System (INIS)

Frolov, Kyrill N.

2003-01-01

In case of a severe accident at a nuclear power plant (NPP) involving the reactor pressure vessel (RPV) melt-through, confident solidification of ex-vessel corium is the imperative condition of its safe retention within the plant containment. The rate-determining process for solidification of ex-vessel coriums in the long-term is the chemical diffusion in the liquid phase at the solid-liquid interface. The process of chemical diffusion in the diffusive boundary layer can evolve taking on different rates, depending on the boundary conditions and the melt composition. Nonetheless, the chemical diffusion rates would entwine the self-diffusivities of corium constituents, which in turn would depend on the melt chemical composition. This work looks at some aspects of analytical and experimental determination of self-diffusivities of corium constituents. Following the corium-concrete interaction, an ex-vessel corium melt would contain several chemical components, including a fraction of silica. Accordingly, ex-vessel corium is considered in this paper as a silicate melts. In the realm of the geological and glass sciences, where silicate melts are most often discussed, the diffusive transport and viscous flow are conceived interrelated from a phenomenological point of view. Though the viscous and diffusive mass transfer mechanisms are not identical for different species even in the same melt, a combination of semi-empirical models can still provide an estimation of the diffusion thresholds in ex-vessel corium melts. Thus, the first part of this paper presents an analysis of the applicability of such empirical models for simple silicate melts based on the published data. This is followed by an estimation of diffusivities in melt compositions typical of ex-vessel coriums. Alternatively, although the general trend towards a coupled description of the viscous flow and diffusion for ex-vessel corium melts seems promising, it is limited to published data on self-diffusivities of

The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

International Nuclear Information System (INIS)

Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

1988-09-01

High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs

Improved Rock Core Sample Break-off, Retention and Ejection System, Phase I

Data.gov (United States)

National Aeronautics and Space Administration — The proposed effort advances the design of an innovative core sampling and acquisition system with improved core break-off, retention and ejection features. The...

Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

Energy Technology Data Exchange (ETDEWEB)

Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

1998-01-01

This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

Round window electrode insertion potentiates retention in the scala tympani.

Science.gov (United States)

Connor, Stephen E J; Holland, N Julian; Agger, Andreas; Leong, Annabelle C; Varghese, Re Ajay; Jiang, Dan; Fitzgerald O'Connor, Alec

2012-09-01

The round window membrane (RWM)-intentioned approach is superior to the traditional bony cochleostomy (BC) approach in obtaining electrode placement within the scala tympani (ST). Cochlear implant outcome is influenced by several factors, including optimal placement and retention of the electrode array within the ST. The present study aimed to assess whether the RWM route is superior to a traditional BC for placement and retention of the electrode array in the ST. This was a prospective consecutive non-randomized comparison study. All patients were implanted with the Advanced Bionics 1J electrode array. The RWM approach (n = 32) was compared with a traditional BC group (n = 33). The outcome measure was the electrode position as judged within the scalar chambers at four points along the basal turn using postoperative computed tomography (CT). When the mean position scores were compared, the RWM-intentioned group had significantly more electrodes directed towards the ST compartment than the BC group (p scala vestibuli.

Progress and Achievements on the R&D Activities for ITER Vacuum Vessel

Energy Technology Data Exchange (ETDEWEB)

Nakahira, M. [Japan Atomic Energy Research Institute (JAERI); Koizumi, K. [Japan Atomic Energy Research Institute (JAERI); Takahashi, H. [Japan Atomic Energy Research Institute (JAERI); Onozuka, M. [ITER Joint Central Team, Garching, Germany; Ioki, K. [ITER Joint Central Team, Garching, Germany; Kuzumin, E. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Krylov, V. [D.V. Efremov Scientific Research Institute, St. Petersburg, Russia; Maslakowski, J. [Oak Ridge National Laboratory (ORNL); Nelson, Brad E [ORNL; Jones, L. [Max-Planck Institute, Garching, Germany; Danner, W. [Max-Planck Institute, Garching, Germany; Maisonnier, D. [Max-Planck Institute, Garching, Germany

2001-01-01

The ITER vacuum vessel (VV) is designed to be large double-walled structure with a D-shaped crosssection. The achievable fabrication tolerance of this structure was unknown due to the size and complexity of shape. The Full-scale Sector Model of ITER Vacuum Vessel, which was 15m in height, was fabricated and tested to obtain the fabrication and assembly tolerances. The model was fabricated within the target tolerance of 5mm and welding deformation during assembly operation was obtained. The port structure was also connected using remotized welding tools to demonstrate the basic maintenance activity. In parallel, the tests of advanced welding, cutting and inspection system were performed to improve the efficiency of fabrication and maintenance of the Vacuum Vessel. These activities show the feasibility of ITER Vacuum Vessel as feasible in a realistic way. This paper describes the major progress, achievement and latest status of the R&D activities on the ITER vacuum vessel.

Improved Rock Core Sample Break-off, Retention and Ejection System, Phase II

Data.gov (United States)

National Aeronautics and Space Administration — The proposed effort advances the design of an innovative core sampling and acquisition system with improved core break-off, retention and ejection features. Phase 1...

Factors influencing recruitment and retention of healthcare workers in rural and remote areas in developed and developing countries: an overview

Directory of Open Access Journals (Sweden)

Gisèle Irène Claudine Mbemba

2016-12-01

Full Text Available Shortage of healthcare workers in rural and remote areas remains a growing concern both in developed and developing countries. This review aims to synthesize the significant factors impacting healthcare professionals’ recruitment and retention in rural and remote areas, and to identify those relevant for developing countries. This paper included the following steps: exploring scientific literature through predetermined criteria and extracting relevant information by two independents reviewers. The AMSTAR tool was used to assess the methodological quality. Of the 224 screened publications, 15 reviews were included. Four reviews focused on recruitment factors, and another four reviews focused on retention factors. The remaining focused both on recruitment and retention factors. The most important factors influencing recruitment were rural background and rural origin, followed by career development. Opportunities for professional advancement, professional support networks and financial incentives were factors impacting retention. While the main factors influencing recruitment and retention have been largely explored in the literature, the evidence on strategies to reduce the shortage of healthcare workers in rural area, particularly in developing countries, is low. Further research in this field is needed.

Proactive life extension of pressure vessels

Science.gov (United States)

Mager, Lloyd

1998-03-01

For a company to maintain its competitive edge in today's global market every opportunity to gain an advantage must be exploited. Many companies are strategically focusing on improved utilization of existing equipment as well as regulatory compliance. Abbott Laboratories is no exception. Pharmaceutical companies such as Abbott Laboratories realize that reliability and availability of their production equipment is critical to be successful and competitive. Abbott Laboratories, like many of our competitors, is working to improve safety, minimize downtime and maximize the productivity and efficiency of key production equipment such as the pressure vessels utilized in our processes. The correct strategy in obtaining these objectives is to perform meaningful inspection with prioritization based on hazard analysis and risk. The inspection data gathered in Abbott Laboratories pressure vessel program allows informed decisions leading to improved process control. The results of the program are reduced risks to the corporation and employees when operating pressure retaining equipment. Accurate and meaningful inspection methods become the cornerstone of a program allowing proper preventative maintenance actions to occur. Successful preventative/predictive maintenance programs must utilize meaningful nondestructive evaluation techniques and inspection methods. Nondestructive examination methods require accurate useful tools that allow rapid inspection for the entire pressure vessel. Results from the examination must allow the owner to prove compliance of all applicable regulatory laws and codes. At Abbott Laboratories the use of advanced NDE techniques, primarily B-scan ultrasonics, has provided us with the proper tools allowing us to obtain our objectives. Abbott Laboratories uses B-scan ultrasonics utilizing a pulse echo pitch catch technique to provide essential data on our pressure vessels. Equipment downtime is reduced because the nondestructive examination usually takes

Advanced ultrasonic and eddy current examinations of the reactor vessel

International Nuclear Information System (INIS)

Cvitanovic, M.; Zado, V.

1996-01-01

In order to improve safety and reliability of nuclear power plant components, the existing examination methods are permanently developed as well as the new methods of examination are implemented. For the same reason, beside referent requirements, complementary NDE methods are utilized. Some examination methods techniques are not required to be used by referent safety codes and standards but they are frequently practiced as additional prevention to the component failure. This article presents the state of the art methods and techniques currently applied for examination of the reactor vessel base material, clad and weld materials. (author)

Faculty Peer Networks: Role and Relevance in Advancing Agency and Gender Equity

Science.gov (United States)

O'Meara, KerryAnn; Stromquist, Nelly P.

2015-01-01

Organisational efforts to alter gender asymmetries are relatively rare, yet they are taking place in a number of universities. In the USA, sponsored by the National Science Foundation, ADVANCE programmes implement a number of interventions to improve the recruitment, retention, and advancement of women faculty. This study focused on one common…

The ultrasonic wave pattern analysis and the frequency diversity signal processing in multi-layered gap measurement for in-vessel corium retention

International Nuclear Information System (INIS)

Koo, K. M.; Kim, J. H.; Kim, S. B.; Kim, H. D.

1999-01-01

A gap between a molten material and a lower head vessel is formed in the LAVA experiment, a phase 1 study of SONATA-IV program. In this paper, the quantitative results of the gap measurement using an off-line ultrasonic pulse echo method by frequency diversity signal processing are presented. However, the gap measurement signal using an ordinary ultrasonic test would be lack of reliability due to the structural complexity of the specimen. The structural complexity may result from the external reason from the shape and the internal reason from the material characteristics. This paper aims at the development of an appropriate ultrasonic test method, by analyzing the problems from the internal characteristic reason. In this test, the signal of the propagational direction and reflectional direction through solid-liquid-solid specimen was analyzed to understand the behavior of the reflectional signal in a multi-layered structure by filling the gap with water between the melt and the lower head vessel

Hypofractionated Image-Guided IMRT in Advanced Pancreatic Cancer With Simultaneous Integrated Boost to Infiltrated Vessels Concomitant With Capecitabine: A Phase I Study

Energy Technology Data Exchange (ETDEWEB)

Passoni, Paolo, E-mail: passoni.paolo@hsr.it [Department of Radiation Oncology, San Raffaele Scientific Institute, Milan (Italy); Reni, Michele [Department of Medical Oncology, San Raffaele Scientific Institute, Milan (Italy); Cattaneo, Giovanni M. [Department of Medical Physics, San Raffaele Scientific Institute, Milan (Italy); Slim, Najla [Department of Radiation Oncology, San Raffaele Scientific Institute, Milan (Italy); Cereda, Stefano [Department of Medical Oncology, San Raffaele Scientific Institute, Milan (Italy); Balzano, Gianpaolo; Castoldi, Renato [Department of Surgery, San Raffaele Scientific Institute, Milan (Italy); Longobardi, Barbara [Department of Medical Physics, San Raffaele Scientific Institute, Milan (Italy); Bettinardi, Valentino; Gianolli, Luigi [Department of Nuclear Medicine, San Raffaele Scientific Institute, Milan (Italy); Gusmini, Simone [Department of Radiology, San Raffaele Scientific Institute, Milan (Italy); Staudacher, Carlo [Department of Surgery, San Raffaele Scientific Institute, Milan (Italy); Calandrino, Riccardo [Department of Medical Physics, San Raffaele Scientific Institute, Milan (Italy); Di Muzio, Nadia [Department of Radiation Oncology, San Raffaele Scientific Institute, Milan (Italy)

2013-12-01

Purpose: To determine the maximum tolerated radiation dose (MTD) of an integrated boost to the tumor subvolume infiltrating vessels, delivered simultaneously with radical dose to the whole tumor and concomitant capecitabine in patients with pretreated advanced pancreatic adenocarcinoma. Methods and Materials: Patients with stage III or IV pancreatic adenocarcinoma without progressive disease after induction chemotherapy were eligible. Patients underwent simulated contrast-enhanced four-dimensional computed tomography and fluorodeoxyglucose-labeled positron emission tomography. Gross tumor volume 1 (GTV1), the tumor, and GTV2, the tumor subvolume 1 cm around the infiltrated vessels, were contoured. GTVs were fused to generate Internal Target Volume (ITV)1 and ITV2. Biological tumor volume (BTV) was fused with ITV1 to create the BTV+Internal Target Volume (ITV) 1. A margin of 5/5/7 mm (7 mm in cranium-caudal) was added to BTV+ITV1 and to ITV2 to create Planning Target Volume (PTV) 1 and PTV2, respectively. Radiation therapy was delivered with tomotherapy. PTV1 received a fixed dose of 44.25 Gy in 15 fractions, and PTV2 received a dose escalation from 48 to 58 Gy as simultaneous integrated boost (SIB) in consecutive groups of at least 3 patients. Concomitant chemotherapy was capecitabine, 1250 mg/m{sup 2} daily. Dose-limiting toxicity (DLT) was defined as any treatment-related G3 nonhematological or G4 hematological toxicity occurring during the treatment or within 90 days from its completion. Results: From June 2005 to February 2010, 25 patients were enrolled. The dose escalation on the SIB was stopped at 58 Gy without reaching the MTD. One patient in the 2{sup nd} dose level (50 Gy) had a DLT: G3 acute gastric ulcer. Three patients had G3 late adverse effects associated with gastric and/or duodenal mucosal damage. All patients received the planned dose of radiation. Conclusions: A dose of 44.25 Gy in 15 fractions to the whole tumor with an SIB of 58 Gy to small

Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

International Nuclear Information System (INIS)

Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

2007-01-01

SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

Long discharge particle balance and fuel retention in Tore Supra

International Nuclear Information System (INIS)

Pegourie, B.; Brosset, C.; Delchambre, E.; Loarer, T.; Tsitrone, E.; Bucalossi, J.; Gunn, J.; Reichle, R.; Khodja, H.; Lafon, C.; Parent, P.

2003-01-01

In the new CIEL configuration of Tore-Supra, all the plasma facing components are actively cooled. The surface of the wall covered with carbon is about 15 m 2 (2-4 m 2 of which in close interaction with the plasma). This configuration allowed to maintain steady-state plasma conditions during up to 4 min 25 s. In these experiments, the required gas injection for maintaining the prescribed density remains constant during the whole discharge. The exhausted flux is also constant and equal to 40-50% of the injected flux. Therefore, 50-60% of the injected particles remain trapped in the vessel, the total retention being proportional to the plasma duration. Since the gas recovered between shots or by conditioning techniques is far from balancing the injected gas, huge quantities of deuterium remain indefinitely trapped in the wall, which appears as an infinite reservoir. This reservoir is believed to be dominated by co-deposited layers, as observed in several places of the vessel. The thickest deposits (up to 800 μm) are observed on the leading edge of the neutralizers of the exhaust system. They display a column-like shape (typical growth rate ∼ 20 nm/s) and have a graphite-like structure. Their deuterium content is modest. Conversely, when coming from regions protected from the direct plasma flux, the deposits show a smoother shape and their deuterium content is typically an order of magnitude larger. (authors)

Earthquake-proof supporting structure in reactor vessel

International Nuclear Information System (INIS)

Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

1990-01-01

Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

Revisiting the reactor pressure vessel for long-time operation

International Nuclear Information System (INIS)

Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

2013-01-01

The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

RETENTION AND PENETRATION OF CCA IN WOOD OF FIRST AND SECOND ROTATION OF Eucalyptus urophylla S.T. Blake

Directory of Open Access Journals (Sweden)

Mara Lúcia Agostini Valle

2013-06-01

Full Text Available http://dx.doi.org/10.5902/198050989292This study aimed to evaluate the retention and penetration of copper chrome arsenate (CCA type C as well as some wood properties of two rotations of two natural hybrid of Eucalyptus urophylla S. T. Blake, aiming their use as treated wood. The study was conducted with material from commercial plantations, with 63 months of age. For the wood characterization, the relationship sapwood-heartwood, the basic density and the size of fibers and vessels were evaluated. For the evaluation of treated wood, the penetration and retention of copper chrome arsenate (CCA type C were determined. Four trees per rotation and genetic material were used, which subsequently were split into three logs, in a total of 12 logs for each treatment. The preservative treatment was performed using the full cell process in autoclave using CCA solution with 2% concentration of active ingredients. The treatment process used was effective under the conditions required by the NBR 9480, with retention values ​​higher than the minimum required by the standard, which is 6.5 kg/m3 of CCA per treated wood, and in addition, provided deep penetration and regular condom in sapwood of all timber treated. There are no restrictions on the use of wood from the first and second rotation for preservative treatment, based on the properties evaluated. There was no correlation between the type C CCA retention and wood properties evaluated.

Investigation of impulsively loaded pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Brown, N.; Cornwell, R.; Hanner, D.; Leichter, H.; Mohr, P.

1963-10-15

Explosion containment vessels for containing from 2,000 to 3,000 five ton nuclear explosions are considered. Analysis methods appear adequate and lowest weights using the most advanced materials available in the next five years are projected.None of these materials can be fabricated today and all require extensive development. Present material technology limits the choice of materials and defines the weight. The addition of safety factors and fixtures (nozzles, etc.) will add to this weight considerably, and may well radically alter the vessel response. Improvements in the strength weight ratios of metals and glasses over those considered in this report do not appear reasonable at this time. Winding schemes to utilize the high strength of steel wires and somehow maintain a reasonable thickness appear to offer the most promise. A `ductile` beryllium would of course offer vast improvement, but no indications that this is being developed have appeared and all presently known beryllium is much too brittle.

Under Water Thermal Cutting of the Moderator Vessel and Thermal Shield

International Nuclear Information System (INIS)

Loeb, A.; Sokcic-Kostic, M.; Eisenmann, B.; Prechtl, E.

2007-01-01

This paper presents the segmentation of the in 8 meter depth of water and for cutting through super alloyed moderator vessel and of the thermal shield of the MZFR stainless steel up to 130 mm wall thickness. Depending on the research reactor by means of under water plasma and contact arc metal cutting. The moderator vessel and the thermal shield are the most essential parts of the MZFR reactor vessel internals. These components have been segmented in 2005 by means of remotely controlled under water cutting utilizing a special manipulator system, a plasma torch and CAMC (Contact Arc Metal Cutting) as cutting tools. The engineered equipment used is a highly advanced design developed in a two years R and D program. It was qualified to cut through steel walls of more than 100 mm thickness in 8 meters water depth. Both the moderator vessel and the thermal shield had to be cut into such size that the segments could afterwards be packed into shielded waste containers each with a volume of roughly 1 m 3 . Segmentation of the moderator vessel and of the thermal shield was performed within 15 months. (author)

Prevention of non-ductile fracture in 6061-T6 aluminum nuclear pressure vessels

International Nuclear Information System (INIS)

Yahr, G.T.

1995-01-01

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Committee has approved rules for the use of 6061-T6 and 6061-T651 aluminum for the construction of Class 1 welded nuclear pressure vessels for temperatures not exceeding 149 C (300 F). Nuclear Code Case N-519 allows the use of this aluminum in the construction of low temperature research reactors such as the Advanced Neutron Source. The rules for protection against non-ductile fracture are discussed. The basis for a value of 25.3 MPa √m (23 ksi √in.) for the critical or reference stress intensity factor for use in the fracture analysis is presented. Requirements for consideration of the effects of neutron irradiation on the fracture toughness are discussed

Tempest in a vessel

International Nuclear Information System (INIS)

Barre, Bertrand

2015-01-01

As the ASN made some statements about anomalies of carbon content in the EPR vessel bottom and top, the author recalls and comments some technical issues to better understand the information published on this topic. He notably addresses the role of the vessel, briefly indicates its operating conditions, shape and structure, and mechanical components for the top, its material and mechanical properties, and test samples used to assess mechanical properties. He also comments the phenomenon of radio-induced embrittlement, the vessel manufacturing process, and evokes the applicable regulations. He quotes and comments statements made by the ASN and Areva which evoke further assessments of the concerned components

Multiple shell pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.

1988-01-01

A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

Uncovering client retention antecedents in service organizations

Directory of Open Access Journals (Sweden)

Mari Jansen van Rensburg

2014-01-01

Full Text Available This paper develops a multi-dimensional model of retention to provide a more complete and integrated view of client retention and its determinants in service contexts. To uncover the antecedents of client retention, social and economic exchanges were reviewed under the fundamental ideas of the Social Exchange Theory. Findings from a survey of senior South African advertising executives suggest that client retention is the result of evaluative as well as relational factors that can influence client responses. Despite contractual obligations, advertisers are willing to pay the costs and make the sacrifices of switching should their expectations be unmet. An important contribution of this study is the use of multi-item scales to measure retention. The model developed provides valuable insight to agencies on client retention management and the optimal allocation of resources for maximum customer equity. This model may also be applied to other service organisations to provide insight to client retention.

Pressurized wet digestion in open vessels (T11)

International Nuclear Information System (INIS)

Kettisch, P.; Maichin, P.; Zischka, M.; Knapp, G.

2002-01-01

Full text: Pressurized wet digestion in closed vessels, microwave assisted or with conventional conductive heating, is the most important sample preparation technique for digestion or leaching procedures in element analysis. In comparison to open vessel digestion closed vessel digestion methods have many advantages, but there is one disadvantage - complex and expensive vessel designs. A new technique - pressurized wet digestion in open vessels - combine the advantages of closed vessel sample digestion with the application of simple and cheap open vessels made of quartz or PFA. The vessels are placed in a high pressure Asher HPA, which is adapted with a Teflon liner and filled partly with water. The analytical results with 30 ml quartz vessels, 22 ml PFA vessels and 1.5 ml PIA auto sampler cups will be shown. In principle every dimensions of vessels can be used. The vessels are loaded with sample material (max. 1.5 g with quartz vessels, max. 0.5 g with PFA vessels and 50 mg with auto sampler cups) and digestion reagent. Afterwards the vessels are simply covered with PTFE stoppers and not sealed. The vessels are transferred into a special adapted HPA and digested at temperatures up to 270 o C. The digestion time is 90 min. and cooling down to room temperature 30 min. The analytical results of CRM's are within the certified values and no cross contamination and losses of volatile elements could be observed. (author)

Best practices in doctoral retention: Mentoring

Directory of Open Access Journals (Sweden)

Judie L. Brill

2014-06-01

Full Text Available  The aim of this critical literature review is to outline best practices in doctoral retention and the successful approach of one university to improve graduation success by providing effective mentorship for faculty and students alike. The focus of this literature review is on distance learning relationships between faculty and doctoral students, regarding retention, persistence, and mentoring models. Key phrases and words used in the search and focusing on mentoring resulted in over 20,000 sources. The search was narrowed to include only doctoral study and mentoring. Research questions of interest were: Why do high attrition rates exist for doctoral students? What are the barriers to retention? What are the benefits of doctoral mentoring? What programs do institutions have in place to reduce attrition? The researchers found a key factor influencing doctoral student retention and success is effective faculty mentorship. In particular, the design of a mentoring and faculty training program to increase retention and provide for success after graduation is important. This research represents a key area of interest in the retention literature, as institutions continue to search for ways to better support students during their doctoral programs and post-graduation. DOI: 10.18870/hlrc.v4i2.186

Retention of gaseous isotopes

International Nuclear Information System (INIS)

Yarbro, O.O.; Mailen, J.C.; Stephenson, M.J.

1977-01-01

Retention of gaseous fission products during fuel reprocessing has, in the past, been limited to a modest retention of 131 I when processing fuels decayed less than about 180 days. The projected rapid growth of the nuclear power industry along with a desire to minimize environmental effects is leading to the reassessment of requirements for retention of gaseous fission products, including 131 I, 129 I, 85 Kr, 3 H, and 14 C. Starting in the late 1960s, a significant part of the LMFBR reprocessing development program has been devoted to understanding the behavior of gaseous fission products in plant process and effluent streams and the development of advanced systems for their removal. Systems for iodine control include methods for evolving up to 99% of the iodine from dissolver solutions to minimize its introduction and distribution throughout downstream equipment. An aqueous scrubbing system (Iodox) using 20 M HNO 3 as the scrubbing media effectively removes all significant iodine forms from off-gas streams while handling the kilogram quantities of iodine present in head-end and dissolver off-gas streams. Silver zeolite is very effective for removing iodine forms at low concentration from the larger-volume plant off-gas streams. Removal of iodine from plant liquid effluents by solid sorbents either prior to or following final vaporization appears feasible. Krypton is effectively released during dissolution and can be removed from the relatively small volume head-end and dissolver off-gas stream. Two methods appear applicable for removal and concentration of krypton: (1) selective absorption in fluorocarbons, and (2) cryogenic absorption in liquid nitrogen. The fluorocarbon absorption process appears to be rather tolerant of the normal contaminants (H 2 O, CO 2 , NOsub(x), and organics) present in typical reprocessing plant off-gas whereas the cryogenic system requires an extensive feed gas pretreatment system. Retention of tritium in a reprocessing plant is

A Nordic approach to impact assessment of accidents with nuclear-propelled vessels

International Nuclear Information System (INIS)

Reistad, O.; Hustveit, S.; Palsson, S.E.; Hoe, S.; Lahtinen, J.

2012-11-01

The MareNuc project has identified the parameters in a graded approach to impact assessment for marine nuclear reactors. The graded approach is founded on the following elements: 1) More detailed understanding of previous accidents in nuclear-propelled vessels (initiating events, accident developments, release fractions), including release mechanisms (radionuclide retention in vessel construction); 2) Bench-marking of release scenarios using modelling tools applied in the Nordic countries, in addition to demonstration of generally accessible and free software developed by the IAEA; 3) Other systematic approaches to safety assessments of vessel port calls, and to the design and maintenance of emergency preparedness systems; More specifically, increased emphasis compared to earlier analysis after the Kursk accident is given to the engineered vessel barriers. Relevant standards from impact assessments for commercial nuclear power plants have been identified, such as from the NUREG series. The Nordic approaches to safety evaluation, impact assessments and emergency preparedness organisation was also reported as part of the project. The Canadian approach for international port calls was carefully reported and assessed as part of the project, and commended for its broad and comprehensive approach to reactor and vessel design for the nationalities involved, to the design and maintenance of emergency preparedness systems, and the well-structured and broad cooperation between civilian and military institutions. This approach goes beyond the current approach in the Nordic countries, also in the case of Norway, which experience regular port calls from allied nuclear navies. The overall result is a broader understanding in the Nordic countries for the importance of the various parameters for impact assessment of releases from marine reactors, and to the design and maintenance of an emergency preparedness organisation without detailed knowledge of the installation in question

A Nordic approach to impact assessment of accidents with nuclear-propelled vessels

Energy Technology Data Exchange (ETDEWEB)

Reistad, O. [Institute for Energy Technology, Kjeller (Norway); Hustveit, S. [Norwegian Radiation Protection Authority, Oesteraes (Norway); Palsson, S.E. [Icelandic Radiation Safety Authority, Reykjavik (Iceland); Hoe, S. [Danish Emergency Management Agency, Birkeroed (Denmark); Lahtinen, J. [STUK, Helsinki (Finland)

2012-11-15

The MareNuc project has identified the parameters in a graded approach to impact assessment for marine nuclear reactors. The graded approach is founded on the following elements: 1) More detailed understanding of previous accidents in nuclear-propelled vessels (initiating events, accident developments, release fractions), including release mechanisms (radionuclide retention in vessel construction); 2) Bench-marking of release scenarios using modelling tools applied in the Nordic countries, in addition to demonstration of generally accessible and free software developed by the IAEA; 3) Other systematic approaches to safety assessments of vessel port calls, and to the design and maintenance of emergency preparedness systems; More specifically, increased emphasis compared to earlier analysis after the Kursk accident is given to the engineered vessel barriers. Relevant standards from impact assessments for commercial nuclear power plants have been identified, such as from the NUREG series. The Nordic approaches to safety evaluation, impact assessments and emergency preparedness organisation was also reported as part of the project. The Canadian approach for international port calls was carefully reported and assessed as part of the project, and commended for its broad and comprehensive approach to reactor and vessel design for the nationalities involved, to the design and maintenance of emergency preparedness systems, and the well-structured and broad cooperation between civilian and military institutions. This approach goes beyond the current approach in the Nordic countries, also in the case of Norway, which experience regular port calls from allied nuclear navies. The overall result is a broader understanding in the Nordic countries for the importance of the various parameters for impact assessment of releases from marine reactors, and to the design and maintenance of an emergency preparedness organisation without detailed knowledge of the installation in question

Test of 6-in.-thick pressure vessels. Series 3: intermediate test vessel V-7

International Nuclear Information System (INIS)

Merkle, J.G.; Robinson, G.C.; Holz, P.P.; Smith, J.E.; Bryan, R.H.

1976-08-01

The test of intermediate test vessel V-7 was a crack-initiation fracture test of a 152-mm-thick (6-in.), 990-mm-OD (39-in.) vessel of ASTM A533, grade B, class 1 steel plate with a sharp outside surface flaw 457 mm (18 in.) long and about 135 mm (5.3 in.) deep. The vessel was heated to 91 0 C (196 0 F) and pressurized hydraulically until leakage through the flaw terminated the test at a peak pressure of 147 MPa (21,350 psi). Fracture toughness data obtained by testing precracked Charpy-V and compact-tension specimens machined from a prolongation of the cylindrical test shell were used in pretest analyses of the flawed vessel. The vessel, as expected, did not burst. Upon depressurization, the ruptured ligament closed so as to maintain static pressure without leakage at about 129 MPa

Vessels in Transit - Web Tool

Data.gov (United States)

Department of Transportation — A web tool that provides real-time information on vessels transiting the Saint Lawrence Seaway. Visitors may sort by order of turn, vessel name, or last location in...

Hydrogen storage in insulated pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

1998-08-01

Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

Fractal Branching in Vascular Trees and Networks by VESsel GENeration Analysis (VESGEN)

Science.gov (United States)

Parsons-Wingerter, Patricia A.

2016-01-01

Vascular patterning offers an informative multi-scale, fractal readout of regulatory signaling by complex molecular pathways. Understanding such molecular crosstalk is important for physiological, pathological and therapeutic research in Space Biology and Astronaut countermeasures. When mapped out and quantified by NASA's innovative VESsel GENeration Analysis (VESGEN) software, remodeling vascular patterns become useful biomarkers that advance out understanding of the response of biology and human health to challenges such as microgravity and radiation in space environments.

A Huge Capital Drop with Compression of Femoral Vessels Associated with Hip Osteoarthritis

Directory of Open Access Journals (Sweden)

Tomoya Takasago

2015-01-01

Full Text Available A capital drop is a type of osteophyte at the inferomedial portion of the femoral head commonly observed in hip osteoarthritis (OA, secondary to developmental dysplasia. Capital drop itself is typically asymptomatic; however, symptoms can appear secondary to impinge against the acetabulum or to irritation of the surrounding tissues, such as nerves, vessels, and tendons. We present here a case of unilateral leg edema in a patient with hip OA, caused by a huge bone mass occurring at the inferomedial portion of the femoral head that compressed the femoral vessels. We diagnosed this bone mass as a capital drop secondary to hip OA after confirming that the mass occurred at least after the age of 63 years based on a previous X-ray. We performed early resection and total hip arthroplasty since the patient’s hip pain was due to both advanced hip OA and compression of the femoral vessels; moreover, we aimed to prevent venous thrombosis secondary to vascular compression considering the advanced age and the potent risk of thrombosis in the patient. A large capital drop should be considered as a cause of vascular compression in cases of unilateral leg edema in OA patients.

A model for structural analysis of nuclear reactor pressure vessel flanges

International Nuclear Information System (INIS)

Oliveira, C.A. de.

1987-01-01

Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

On the optimal design of risk retention in securitisation

Directory of Open Access Journals (Sweden)

Metin Kaptan

2011-09-01

Full Text Available This paper examines the optimal design of retention in securitisation, in order to maximize welfare of screening per unit of retention, assuming that screening is costly and that the bank intends to securitise its loans. In contrast to the focus of previous literature on tranche retention, we deviate from the constitutional mechanisms of tranche retention to present a pareto-optimal method of tranche retention. Unlike the current ad-hoc-regulations, we derive the optimal design of retention from a utility maximization problem. We show that the level of retention per tranche should be dependent on the rate of credit default, i.e. the higher the rate of default, the higher the optimal rate of retention required to provide an incentive to screen carefully. From this approach, it follows that the rate of retention per tranche should be higher, the higher the position within the ranking order of subordination. Accordingly, the efficiency of tranche retention can be enhanced, reducing the level of retention required to maintain a given level of screening-effort. This retention design entails a recovery of the bank’s equity capital, thereby increasing liquidity and lending capacities.

Evaluation of fatigue damage of pressure vessel materials by observation of microstructures

International Nuclear Information System (INIS)

Yoshida, Kazuo

1994-01-01

As the important factor as the secular change mode of pressure vessel materials, there is fatigue damage. In USA, there is the move to use LWRs by extending their life, and it becomes necessary to show the soundness of the structures of machinery and equipment for long period. For exactly evaluating the soundness of the structures of machinery and equipment, it is important to clarify the degree of secular deterioration of the materials. In this report, by limiting to the fatigue damage of LWR pressure vessel steel, the method of grasping the change of microstructure and the method of estimating the degree of fatigue damage from the change of microstructure are shown. The change of microstructure arising in materials due to fatigue advances in the following steps, namely, the multiplication of dislocations, the tangling of dislocations, the formation of cell structure, the turning of cells, the formation of microcracks, the growth of cracks and fracture. In the case of pressure vessel steel, due to the quenching and tempering, the cell structure is formed from the beginning, and the advance of fatigue is recognized as the increase of the turning angle of cell structures. The detection of fatigue damage by microstructure is reported. (K.I.)

Vessel size measurements in angiograms: Manual measurements

International Nuclear Information System (INIS)

Hoffmann, Kenneth R.; Dmochowski, Jacek; Nazareth, Daryl P.; Miskolczi, Laszlo; Nemes, Balazs; Gopal, Anant; Wang Zhou; Rudin, Stephen; Bednarek, Daniel R.

2003-01-01

Vessel size measurement is perhaps the most often performed quantitative analysis in diagnostic and interventional angiography. Although automated vessel sizing techniques are generally considered to have good accuracy and precision, we have observed that clinicians rarely use these techniques in standard clinical practice, choosing to indicate the edges of vessels and catheters to determine sizes and calibrate magnifications, i.e., manual measurements. Thus, we undertook an investigation of the accuracy and precision of vessel sizes calculated from manually indicated edges of vessels. Manual measurements were performed by three neuroradiologists and three physicists. Vessel sizes ranged from 0.1-3.0 mm in simulation studies and 0.3-6.4 mm in phantom studies. Simulation resolution functions had full-widths-at-half-maximum (FWHM) ranging from 0.0 to 0.5 mm. Phantom studies were performed with 4.5 in., 6 in., 9 in., and 12 in. image intensifier modes, magnification factor = 1, with and without zooming. The accuracy and reproducibility of the measurements ranged from 0.1 to 0.2 mm, depending on vessel size, resolution, and pixel size, and zoom. These results indicate that manual measurements may have accuracies comparable to automated techniques for vessels with sizes greater than 1 mm, but that automated techniques which take into account the resolution function should be used for vessels with sizes smaller than 1 mm

Preliminary Design of Optimized Reactor Insulator for Severe Accident Mitigation of APR1400

International Nuclear Information System (INIS)

Heo, Sun; Lee, Jae-Gon; Kang, Yong-Chul

2007-01-01

APR1400, a Korean evolutionary advance light water reactor, has many advanced safety feature to prevent and mitigate of design basis accident (DBA) and severe accident. When reactor cooling system (RCS) fails to cooling its core, the core melted down and the molten core gathers together on bottom of reactor vessel. The molten core hurts reactor vessel and is released to containment, which raises the release of radioactive isotopes and the heating of the containment atmosphere. Finally, the corium is accumulated in the bottom of reactor cavity and it also raises the Molten Core and Concrete Interaction (MCCI) and the heating of containment atmosphere. There are two strategies to cooling molten core. Those are in-vessel retention and ex-vessel cooling. At the early stage of APR1400 design, only ex-vessel cooling which is cooling of the molten core outside the vessel after vessel failure is considered based on EPRI Utility Requirement Document (URD) for Evolutionary LWR. However, a need has been arisen to reflect current research findings on severe accident phenomena and mitigation technologies to Korean URD and IVRERVC (In-Vessel corium Retention using Ex-Reactor Vessel Cooling) was adopted APR1400. The ERVC is not considered as a licensing design basis but based on the defense-in-depth principle and safety margin basis, which is the top-tier requirement of the severe accident mitigation design as stated in the KURD. The Severe Accident Management strategy for APR1400 is intended to aid the plant operating staff to secure reactor vessel integrity in the early stage of the severe accident. As a part of a design implementation of IVR-ERVC for APR1400, we developed the preliminary design requirement, design specification and conceptual design

Effects of sodium polyacrylate on water retention and infiltration capacity of a sandy soil

OpenAIRE

Zhuang, Wenhua; Li, Longguo; Liu, Chao

2013-01-01

Based on the laboratory study, the effects of sodium polyacrylate (SP) was investigated at 5 rates of 0, 0.08, 0.2, 0.5, and 1%, on water retention, saturated hydraulic conductivity(Ks), infiltration characteristic and water distribution profiles of a sandy soil. The results showed that water retention and available water capacity effectively increased with increasing SP rate. The Ks and the rate of wetting front advance and infiltration under certain pond infiltration was significantly reduc...

Debulking surgery for elephantiasis nostras with large ectatic podoplanin-negative lymphatic vessels in patients with lipo-lymphedema.

Science.gov (United States)

Wollina, Uwe; Heinig, Birgit; Schönlebe, Jaqueline; Nowak, Andreas

2014-01-01

Elephantiasis nostras is a rare complication in advanced lipo-lymphedema. While lipedema can be treated by liposuction and lymphedema by decongestive lymphatic therapy, elephantiasis nostras may need debulking surgery. We present 2 cases of advanced lipo-lymphedema complicated by elephantiasis nostras. After tumescent microcannular laser-assisted liposuction both patients underwent a debulking surgery with a modification of Auchincloss-Kim's technique. Histologic examination of the tissue specimen was performed. The surgical treatment was well tolerated and primary healing was uneventful. After primary wound healing and ambulation of the patients, a delayed ulceration with lymphorrhea developed. It was treated by surgical necrectomy and vacuum-assisted closure leading to complete healing. Mobility of the leg was much improved. Histologic examination revealed massive ectatic lymphatic vessels nonreactive for podoplanin. Debulking surgery can be an adjuvant technique for elephantiasis nostras in advanced lipo-lymphedema. Although delayed postoperative wound healing problems were observed, necrectomy and vacuum-assisted closure achieved a complete healing. Histologic data suggest that the ectatic lymphatic vessels in these patients resemble finding in podoplanin knockout mice. The findings would explain the limitations of decongestive lymphatic therapy and tumescent liposuction in such patients and their predisposition to relapsing erysipelas.

Novel word retention in bilingual and monolingual speakers

Directory of Open Access Journals (Sweden)

Pui Fong eKan

2014-09-01

Full Text Available The goal of this research was to examine word retention in bilinguals and monolinguals. Long-term word retention is an essential part of vocabulary learning. Previous studies have documented that bilinguals outperform monolinguals in terms of retrieving newly-exposed words. Yet, little is known about whether or to what extent bilinguals are different from monolinguals in word retention. Participants were 30 English-speaking monolingual adults and 30 bilingual adults who speak Spanish as a home language and learned English as a second language during childhood. In a previous study (Kan, Sadagopan, Janich, & Andrade, 2014, the participants were exposed to the target novel words in English, Spanish, and Cantonese. In this current study, word retention was measured a week after the fast mapping task. No exposures were given during the one-week interval. Results showed that bilinguals and monolinguals retain a similar number of words. However, participants produced more words in English than in either Spanish or Cantonese. Correlation analyses revealed that language knowledge plays a role in the relationships between fast mapping and word retention. Specifically, within- and across-language relationships between bilinguals’ fast mapping and word retention were found in Spanish and English, by contrast, within-language relationships between monolinguals’ fast mapping and word retention were found in English and across-language relationships between their fast mapping and word retention performance in English and Cantonese. Similarly, bilinguals differed from monolinguals in the relationships among the word retention scores in three languages. Significant correlations were found among bilinguals’ retention scores. However, no such correlations were found among monolinguals’ retention scores. The overall findings suggest that bilinguals’ language experience and language knowledge most likely contribute to how they learn and retain new words.

Novel word retention in bilingual and monolingual speakers.

Science.gov (United States)

Kan, Pui Fong; Sadagopan, Neeraja

2014-01-01

The goal of this research was to examine word retention in bilinguals and monolinguals. Long-term word retention is an essential part of vocabulary learning. Previous studies have documented that bilinguals outperform monolinguals in terms of retrieving newly-exposed words. Yet, little is known about whether or to what extent bilinguals are different from monolinguals in word retention. Participants were 30 English-speaking monolingual adults and 30 bilingual adults who speak Spanish as a home language and learned English as a second language during childhood. In a previous study (Kan et al., 2014), the participants were exposed to the target novel words in English, Spanish, and Cantonese. In this current study, word retention was measured a week after the fast mapping task. No exposures were given during the one-week interval. Results showed that bilinguals and monolinguals retain a similar number of words. However, participants produced more words in English than in either Spanish or Cantonese. Correlation analyses revealed that language knowledge plays a role in the relationships between fast mapping and word retention. Specifically, within- and across-language relationships between bilinguals' fast mapping and word retention were found in Spanish and English, by contrast, within-language relationships between monolinguals' fast mapping and word retention were found in English and across-language relationships between their fast mapping and word retention performance in English and Cantonese. Similarly, bilinguals differed from monolinguals in the relationships among the word retention scores in three languages. Significant correlations were found among bilinguals' retention scores. However, no such correlations were found among monolinguals' retention scores. The overall findings suggest that bilinguals' language experience and language knowledge most likely contribute to how they learn and retain new words.

Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

Energy Technology Data Exchange (ETDEWEB)

Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D. [UCSF Benioff Children' s Hospital, Department of Radiology and Biomedical Imaging, San Francisco, CA (United States)

2014-07-15

Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

Computed tomography depiction of small pediatric vessels with model-based iterative reconstruction

International Nuclear Information System (INIS)

Koc, Gonca; Courtier, Jesse L.; Phelps, Andrew; Marcovici, Peter A.; MacKenzie, John D.

2014-01-01

Computed tomography (CT) is extremely important in characterizing blood vessel anatomy and vascular lesions in children. Recent advances in CT reconstruction technology hold promise for improved image quality and also reductions in radiation dose. This report evaluates potential improvements in image quality for the depiction of small pediatric vessels with model-based iterative reconstruction (Veo trademark), a technique developed to improve image quality and reduce noise. To evaluate Veo trademark as an improved method when compared to adaptive statistical iterative reconstruction (ASIR trademark) for the depiction of small vessels on pediatric CT. Seventeen patients (mean age: 3.4 years, range: 2 days to 10.0 years; 6 girls, 11 boys) underwent contrast-enhanced CT examinations of the chest and abdomen in this HIPAA compliant and institutional review board approved study. Raw data were reconstructed into separate image datasets using Veo trademark and ASIR trademark algorithms (GE Medical Systems, Milwaukee, WI). Four blinded radiologists subjectively evaluated image quality. The pulmonary, hepatic, splenic and renal arteries were evaluated for the length and number of branches depicted. Datasets were compared with parametric and non-parametric statistical tests. Readers stated a preference for Veo trademark over ASIR trademark images when subjectively evaluating image quality criteria for vessel definition, image noise and resolution of small anatomical structures. The mean image noise in the aorta and fat was significantly less for Veo trademark vs. ASIR trademark reconstructed images. Quantitative measurements of mean vessel lengths and number of branches vessels delineated were significantly different for Veo trademark and ASIR trademark images. Veo trademark consistently showed more of the vessel anatomy: longer vessel length and more branching vessels. When compared to the more established adaptive statistical iterative reconstruction algorithm, model

Enhanced post wash retention of combed DNA molecules by varying multiple combing parameters.

Science.gov (United States)

Yadav, Hemendra; Sharma, Pulkit

2017-11-01

Recent advances in genomics have created a need for efficient techniques for deciphering information hidden in various genomes. Single molecule analysis is one such technique to understand molecular processes at single molecule level. Fiber- FISH performed with the help of DNA combing can help us in understanding genetic rearrangements and changes in genome at single DNA molecule level. For performing Fiber-FISH we need high retention of combed DNA molecules post wash as Fiber-FISH requires profuse washing. We optimized combing process involving combing solution, method of DNA mounting on glass slides and coating of glass slides to enhance post-wash retention of DNA molecules. It was found that average number of DNA molecules observed post-wash per field of view was maximum with our optimized combing solution. APTES coated glass slides showed lesser retention than PEI surface but fluorescent intensity was higher in case of APTES coated surface. Capillary method used to mount DNA on glass slides also showed lesser retention but straight DNA molecules were observed as compared to force flow method. Copyright © 2017 Elsevier Inc. All rights reserved.

Nitrogen Saturation in Highly Retentive Watersheds?

Science.gov (United States)

Daley, M. L.; McDowell, W. H.

2009-12-01

Watershed managers are often concerned with minimizing the amount of N delivered to N-limited estuaries and coastal zones. A major concern is that watersheds might reach N saturation, in which N delivered to coastal zones increases due to declines in the efficiency of N retention despite constant or even reduced N inputs. We have quantified long-term changes in N inputs (atmospheric deposition, imported food and agricultural fertilizers), outputs (N concentration and export) and retention in the urbanizing Lamprey River watershed in coastal NH. Overall, the Lamprey watershed is 70% forested, receives about 13.5 kg N/ha/yr and has a high rate of annual N retention (85%). Atmospheric deposition (8.7 kg/ha/yr) is the largest N input to the watershed. Of the 2.2 kg N/ha/yr exported in the Lamprey River, dissolved organic N (DON) is the dominant form (50% of total) and it varies spatially throughout the watershed with wetland cover. Nitrate accounts for 30% of the N exported, shows a statistically significant increase from 1999 to 2009, and its spatial variability in both concentration and export is related to human population density. In sub-basins throughout the Lamprey, inorganic N retention is high (85-99%), but the efficiency of N retention declines sharply with increased human population density and associated anthropogenic N inputs. N assimilation in the vegetation, denitrification to the atmosphere and storage in the groundwater pool could all be important contributors to the current high rates of N retention. The temporal and spatial patterns that we have observed in nitrate concentration and export are driven by increases in N inputs and impervious surfaces over time, but the declining efficiency of N retention suggests that the watershed may also be reaching N saturation. The downstream receiving estuary, Great Bay, already suffers from low dissolved oxygen levels and eelgrass loss in part due to N loading from the Lamprey watershed. Targeting and reducing

Novel word retention in sequential bilingual children.

Science.gov (United States)

Kan, Pui Fong

2014-03-01

Children's ability to learn and retain new words is fundamental to their vocabulary development. This study examined word retention in children learning a home language (L1) from birth and a second language (L2) in preschool settings. Participants were presented with sixteen novel words in L1 and in L2 and were tested for retention after either a 2-month or a 4-month delay. Results showed that children retained more words in L1 than in L2 for both of the retention interval conditions. In addition, children's word retention was associated with their existing language knowledge and their fast-mapping performance within and across language. The patterns of association, however, were different between L1 and L2. These findings suggest that children's word retention might be related to the interactions of various components that are operating within a dynamic system.

Life extension of the St. Lucie unit 1 reactor vessel

International Nuclear Information System (INIS)

Rowan, G.A.; Sun, J.B.; Mott, S.L.

1991-01-01

In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

Evaluation of thermal ratcheting of reactor vessel wall near the sodium surface

International Nuclear Information System (INIS)

Take, Kohji; Fujioka, Terutaka; Yano, Kazutaka

1989-01-01

Plastic ratcheting of reactor vessels may occur by an axially moving thermal gradient without primary stress. So there is a need to establish a proper prediction method for the plastic ratcheting. In this study, inelastic FEM analyses of reactor vessel model by using an advanced constitutive equation were carried out in order to comprehend plastic ratcheting behaviour of cylinder which subject to an axially moving thermal gradient. As a result of analyses, a basic mechanism of this ratcheting was found. And it also indicated that cyclic hardening behaviour will became important for development of evaluation method. (author)

Deuterium retention in molten salt electrodeposition tungsten coatings

International Nuclear Information System (INIS)

Zhou, Hai-Shan; Xu, Yu-Ping; Sun, Ning-Bo; Zhang, Ying-Chun; Oya, Yasuhisa; Zhao, Ming-Zhong; Mao, Hong-Min; Ding, Fang; Liu, Feng; Luo, Guang-Nan

2016-01-01

Highlights: • We investigate D retention in electrodeposition W coatings. • W coatings are exposed to D plasmas in the EAST tokamak. • A cathodic current density dependence on D retention is found. • Electrodeposition W exhibits lower D retention than VPS-W. - Abstract: Molten salt electrodeposition is a promising technology to manufacture the first wall of a fusion reactor. Deuterium (D) retention behavior in molten salt electrodeposition tungsten (W) coatings has been investigated by D-plasma exposure in the EAST tokamak and D-ion implantation in an ion beam facility. Tokamak exposure experiments demonstrate that coatings prepared with lower current density exhibit less D retention and milder surface damage. Deuterium-ion implantation experiments indicate the D retention in the molten salt electrodeposition W is less than that in vacuum plasma spraying W and polycrystalline W.

Deuterium retention in molten salt electrodeposition tungsten coatings

Energy Technology Data Exchange (ETDEWEB)

Zhou, Hai-Shan [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Xu, Yu-Ping [Science Island Branch of Graduate School, University of Science and Technology of China, Hefei (China); Sun, Ning-Bo; Zhang, Ying-Chun [School of Materials Science and Engineering, University of Science and Technology Beijing, Beijing (China); Oya, Yasuhisa [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka (Japan); Zhao, Ming-Zhong [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Mao, Hong-Min [Science Island Branch of Graduate School, University of Science and Technology of China, Hefei (China); Ding, Fang; Liu, Feng [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Luo, Guang-Nan, E-mail: gnluo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Science Island Branch of Graduate School, University of Science and Technology of China, Hefei (China); Hefei Center for Physical Science and Technology, Hefei (China); Hefei Science Center of Chinese Academy of Science, Hefei (China)

2016-12-15

Highlights: • We investigate D retention in electrodeposition W coatings. • W coatings are exposed to D plasmas in the EAST tokamak. • A cathodic current density dependence on D retention is found. • Electrodeposition W exhibits lower D retention than VPS-W. - Abstract: Molten salt electrodeposition is a promising technology to manufacture the first wall of a fusion reactor. Deuterium (D) retention behavior in molten salt electrodeposition tungsten (W) coatings has been investigated by D-plasma exposure in the EAST tokamak and D-ion implantation in an ion beam facility. Tokamak exposure experiments demonstrate that coatings prepared with lower current density exhibit less D retention and milder surface damage. Deuterium-ion implantation experiments indicate the D retention in the molten salt electrodeposition W is less than that in vacuum plasma spraying W and polycrystalline W.

Analysis and optimization on in-vessel inspection robotic system for EAST

International Nuclear Information System (INIS)

Zhang, Weijun; Zhou, Zeyu; Yuan, Jianjun; Du, Liang; Mao, Ziming

2015-01-01

Since China has successfully built her first Experimental Advanced Superconducting TOKAMAK (EAST) several years ago, great interest and demand have been increasing in robotic in-vessel inspection/operation systems, by which an observation of in-vessel physical phenomenon, collection of visual information, 3D mapping and localization, even maintenance are to be possible. However, it has been raising many challenges to implement a practical and robust robotic system, due to a lot of complex constraints and expectations, e.g., high remanent working temperature (100 °C) and vacuum (10"−"3 pa) environment even in the rest interval between plasma discharge experiments, close-up and precise inspection, operation efficiency, besides a general kinematic requirement of D shape irregular vessel. In this paper we propose an upgraded robotic system with redundant degrees of freedom (DOF) manipulator combined with a binocular vision system at the tip and a virtual reality system. A comprehensive comparison and discussion are given on the necessity and main function of the binocular vision system, path planning for inspection, fast localization, inspection efficiency and success rate in time, optimization of kinematic configuration, and the possibility of underactuated mechanism. A detailed design, implementation, and experiments of the binocular vision system together with the recent development progress of the whole robotic system are reported in the later part of the paper, while, future work and expectation are described in the end.

An automated vessel segmentation of retinal images using multiscale vesselness

International Nuclear Information System (INIS)

Ben Abdallah, M.; Malek, J.; Tourki, R.; Krissian, K.

2011-01-01

The ocular fundus image can provide information on pathological changes caused by local ocular diseases and early signs of certain systemic diseases, such as diabetes and hypertension. Automated analysis and interpretation of fundus images has become a necessary and important diagnostic procedure in ophthalmology. The extraction of blood vessels from retinal images is an important and challenging task in medical analysis and diagnosis. In this paper, we introduce an implementation of the anisotropic diffusion which allows reducing the noise and better preserving small structures like vessels in 2D images. A vessel detection filter, based on a multi-scale vesselness function, is then applied to enhance vascular structures.

Probabilistic retinal vessel segmentation

Science.gov (United States)

Wu, Chang-Hua; Agam, Gady

2007-03-01

Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

Factors affecting the retention of Generation X public servants: An exploratory study

Directory of Open Access Journals (Sweden)

Vhutshilo Masibigiri

2011-03-01

Purpose: The purpose of this study was to explore the factors that affect the retention of Generation X public servants. Motivation for the study: Given their unique characteristics, it is a challenge to retain Generation X employees. This problem may be worse in the public sector than in the private sector, as there are fewer financial rewards in the public service than in the private sector. Research design: The interpretivist paradigm is appropriate for this study. It used a qualitative, empirical approach. The researchers obtained the data through purposive sampling and interviews. Main findings: The study showed that the factors affecting the retention of Generation X public servants include work content, utilisation of skills, career advancement, work–life balance, compensation, security needs, leadership and drive. Practical/managerial implications: Employers, like the civil service, can be proactive in retaining Generation X employees because of the factors that affect their retention. Managers can prevent further pressure on service delivery that the skills shortage has caused if they use the skills the employees already have. Contribution: The article fills a gap, as there has been little research on staff retention. This is particularly true of Generation X employees in South Africa. This article adds information that will improve retention strategies for Generation X employees, particularly in the public service.

Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

International Nuclear Information System (INIS)

Bodmann, E.

1984-01-01

Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

33 CFR 90.3 - Pushing vessel and vessel being pushed: Composite unit.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Pushing vessel and vessel being... HOMELAND SECURITY INLAND NAVIGATION RULES INLAND RULES: INTERPRETATIVE RULES § 90.3 Pushing vessel and vessel being pushed: Composite unit. Rule 24(b) of the Inland Rules states that when a pushing vessel and...

33 CFR 82.3 - Pushing vessel and vessel being pushed: Composite unit.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Pushing vessel and vessel being... HOMELAND SECURITY INTERNATIONAL NAVIGATION RULES 72 COLREGS: INTERPRETATIVE RULES § 82.3 Pushing vessel and vessel being pushed: Composite unit. Rule 24(b) of the 72 COLREGS states that when a pushing vessel and a...

Manufacturing and assembly of the plasma- and outer vessel of the cryostat for Wendelstein 7-X

Energy Technology Data Exchange (ETDEWEB)

Hein, Bernd, E-mail: Bernd.Hein@ipp.mpg.de [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstrasse 1, D-17491 Greifswald (Germany); Cardella, Antonio; Hermann, Dieter; Hansen, Andreas [Max-Planck Institut fuer Plasmaphysik, EURATOM Association, Teilinstitut Greifswald, Wendelsteinstrasse 1, D-17491 Greifswald (Germany); Leher, Franz; Binni, Andreas; Segl, Juergen [MAN Diesel and Turbo SE Deggendorf, Werftstrasse 17, D-94469 Deggendorf (Germany)

2012-02-15

Wendelstein 7-X is an advanced helical stellarator, which is presently under construction at the Greifswald branch of IPP. A set of 70 superconducting coils arranged in five modules provides a twisted shaped magnetic cage for the plasma and allows steady state operation. Operation of the magnet system at cryogenic temperatures requires a cryostat which provides thermal protection and gives access to the plasma. The main components of the cryostat are the plasma vessel, the outer vessel, the ports, and the thermal insulation. The German company, MAN Diesel and Turbo SE Deggendorf (former MAN DWE GmbH Deggendorf), is responsible for the manufacture and assembly of the plasma vessel, the outer vessel and the thermal insulation. This paper describes the manufacturing and assembly technology of the plasma and outer vessel of the cryostat for Wendelstein 7-X.

Manufacturing and assembly of the plasma- and outer vessel of the cryostat for Wendelstein 7-X

International Nuclear Information System (INIS)

Hein, Bernd; Cardella, Antonio; Hermann, Dieter; Hansen, Andreas; Leher, Franz; Binni, Andreas; Segl, Jürgen

2012-01-01

Wendelstein 7-X is an advanced helical stellarator, which is presently under construction at the Greifswald branch of IPP. A set of 70 superconducting coils arranged in five modules provides a twisted shaped magnetic cage for the plasma and allows steady state operation. Operation of the magnet system at cryogenic temperatures requires a cryostat which provides thermal protection and gives access to the plasma. The main components of the cryostat are the plasma vessel, the outer vessel, the ports, and the thermal insulation. The German company, MAN Diesel and Turbo SE Deggendorf (former MAN DWE GmbH Deggendorf), is responsible for the manufacture and assembly of the plasma vessel, the outer vessel and the thermal insulation. This paper describes the manufacturing and assembly technology of the plasma and outer vessel of the cryostat for Wendelstein 7-X.

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Vessel discoloration detection in malarial retinopathy

Science.gov (United States)

Agurto, C.; Nemeth, S.; Barriga, S.; Soliz, P.; MacCormick, I.; Taylor, T.; Harding, S.; Lewallen, S.; Joshi, V.

2016-03-01

Cerebral malaria (CM) is a life-threatening clinical syndrome associated with malarial infection. It affects approximately 200 million people, mostly sub-Saharan African children under five years of age. Malarial retinopathy (MR) is a condition in which lesions such as whitening and vessel discoloration that are highly specific to CM appear in the retina. Other unrelated diseases can present with symptoms similar to CM, therefore the exact nature of the clinical symptoms must be ascertained in order to avoid misdiagnosis, which can lead to inappropriate treatment and, potentially, death. In this paper we outline the first system to detect the presence of discolored vessels associated with MR as a means to improve the CM diagnosis. We modified and improved our previous vessel segmentation algorithm by incorporating the `a' channel of the CIELab color space and noise reduction. We then divided the segmented vasculature into vessel segments and extracted features at the wall and in the centerline of the segment. Finally, we used a regression classifier to sort the segments into discolored and not-discolored vessel classes. By counting the abnormal vessel segments in each image, we were able to divide the analyzed images into two groups: normal and presence of vessel discoloration due to MR. We achieved an accuracy of 85% with sensitivity of 94% and specificity of 67%. In clinical practice, this algorithm would be combined with other MR retinal pathology detection algorithms. Therefore, a high specificity can be achieved. By choosing a different operating point in the ROC curve, our system achieved sensitivity of 67% with specificity of 100%.

Location and repair of air leaks in the ATF vacuum vessel

International Nuclear Information System (INIS)

Schwenterly, S.W.; Gabbard, W.A.; Schaich, C.R.; Yarber, J.L.

1989-01-01

On the basis of partial pressure rate-of-rise and base pressure measurements, it was determined that the Advanced Toroidal Facility (ATF) vacuum vessel had an air leak in the low 10 -4 mbarx ell/s range. Pinpointing this leak by conventional helium leak-checking procedures was not possible, because large portions of the outside of the vessel are covered by the helical field coils and a structural shell. Various alternative leak detection schemes that were considered are summarized and their advantages and disadvantages noted. In the method ultimately employed, gum-rubber patches of various sizes ranging from 12.7 by 12.7 cm to 20.3 by 30.5 cm were positioned on the inside surfaces of the vessel and evacuated by the leak detector (LD). After roughly 5% of the surface was inspected in this way, a leak of >10 -5 mbar xL/s was discovered and localized to an area of 5 by 5 cm. Dye penetrant applied to this area disclosed three pinholes. Two small slag pockets were discovered while these points were being ground out. After these were rewelded, no further leakage could be found in the repaired area. Global leak rates measured after the machine was reevacuated indicated that this leak was about 30% of the overall leak rate. 1 ref., 5 figs., 1 tab

Advanced accumulator for PWR

International Nuclear Information System (INIS)

Ichimura, Taiki; Chikahata, Hideyuki

1997-01-01

Advanced accumulators have been incorporated into the APWR design in order to simplify the safety system configuration and to improve reliability. The advanced accumulators refill the reactor vessel with a large discharge flow rate in a large LOCA, then switch to a small flow rate to continue safety injection for core reflooding. The functions of the conventional accumulator and the low head safety injection pump are integrated into this advanced accumulator. Injection performance tests simulating LOCA conditions and visualization tests for new designs have been carried out. This paper describes the APWR ECCS configuration, the advanced accumulator design and some of the injection performance and visualization test results. It was verified that the flow resistance of the advanced accumulator is independent of the model scale. The similarity law and performance data of the advanced accumulator for applying APWR was established. (author)

African Retentions in Blues and Jazz.

Science.gov (United States)

Meadows, Eddie S.

1979-01-01

The perseverance of African musical characteristics among American Blacks is an historic reality. African retentions have been recorded in Black music of the antebellum period. Various African scales and rhythms permeate Black American music today as evidenced in the retentions found in blues and jazz. (RLV)

Clinical trial management of participant recruitment, enrollment, engagement, and retention in the SMART study using a Marketing and Information Technology (MARKIT) model

Science.gov (United States)

Gupta, Anjali; Calfas, Karen J.; Marshall, Simon J.; Robinson, Thomas N.; Rock, Cheryl L.; Huang, Jeannie S.; Epstein-Corbin, Melanie; Servetas, Christina; Donohue, Michael C.; Norman, Gregory J.; Raab, Fredric; Merchant, Gina; Fowler, James H.; Griswold, William G.; Fogg, B.J.; Patrick, Kevin

2015-01-01

Advances in information technology and near ubiquity of the Internet have spawned novel modes of communication and unprecedented insights into human behavior via the digital footprint. Health behavior randomized controlled trials (RCTs), especially technology-based, can leverage these advances to improve the overall clinical trials management process and benefit from improvements at every stage, from recruitment and enrollment to engagement and retention. In this paper, we report the results for recruitment and retention of participants in the SMART study and introduce a new model for clinical trials management that is a result of interdisciplinary team science. The MARKIT model brings together best practices from information technology, marketing, and clinical research into a single framework to maximize efforts for recruitment, enrollment, engagement, and retention of participants into a RCT. These practices may have contributed to the study’s on-time recruitment that was within budget, 86% retention at 24 months, and a minimum of 57% engagement with the intervention over the 2-year RCT. Use of technology in combination with marketing practices may enable investigators to reach a larger and more diverse community of participants to take part in technology-based clinical trials, help maximize limited resources, and lead to more cost-effective and efficient clinical trial management of study participants as modes of communication evolve among the target population of participants. PMID:25866383

A study of the external cooling capability for the prevention of reactor vessel failure

Energy Technology Data Exchange (ETDEWEB)

Chang, S H; Baek, W P; Moon, S K; Yang, S H; Kim, S H [Korea Advanced Institute of Science Technology, Daejeon (Korea, Republic of)

1994-07-15

This study (a 3-year program) aims to perform a comprehensive assessment of the feasibility of external vessel flooding with respect to advanced pressurized water reactor plants to be built in Korea. During the first year, review of the relevant phenomena and preliminary assessment of the concept have been performed. Also performed is a review of heat transfer correlations for the computer program that will be developed for assessment of the cooling capability of external vessel flooding. Important phenomena that determine the cooling capability of external vessel flooding are (a) the initial transient before formation of molten corium pool, (b) natural convection of in-vessel molten corium pool, (c) radiative heat exchange between the molten corium pool and the upper vessel structures, (d) thermal hydraulics outside the vessel, (e) structural integrity consideration, and (f) long-term phenomena. The adoption of the concept should be decided by considering several factors such as (a) vessel submergence procedure, (b) cooling requirements, (c) vessel design features, (d) steam production, (e) instrumentation needs, and (f) an overall accident management strategy. The external vessel cooling concept looks to be promising. However, further study is required for a reliable decision making. Several correlations are available for the prediction of cooling capability of the present concept. However, it is difficult to define a sufficiently reliable set of correlations; sensitivity studies would be required in assessing the cooling capability with the computer program.

Black Student Retention in Higher Education.

Science.gov (United States)

Lang, Marvel, Ed.; Ford, Clinita A., Ed.

This collection focuses on problems in the recruitment, enrollment and retention of Blacks in higher education in America. The following chapters are provided: "The Black Student Retention Problem in Higher Education: Some Introductory Perspectives" (Marvel Lang); "Early Acceptance and Institutional Linkages in a Model Program of Recruitment,…

Retention and Advancement in the Retail Industry: A Career Ladder Approach.

Science.gov (United States)

Prince, Heath J.

Retailing is the largest industry in the United States, employing roughly 18 percent of the total labor force. However, high turnover resulting from low wages in entry-level positions and the perceptions of retail workers that job security is far from certain and that advancement potential is limited have resulted in low levels of employee…

FFTF and CRBRP reactor vessels

International Nuclear Information System (INIS)

Morgan, R.E.

1977-01-01

The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

Tumor Blood Vessel Dynamics

Science.gov (United States)

Munn, Lance

2009-11-01

``Normalization'' of tumor blood vessels has shown promise to improve the efficacy of chemotherapeutics. In theory, anti-angiogenic drugs targeting endothelial VEGF signaling can improve vessel network structure and function, enhancing the transport of subsequent cytotoxic drugs to cancer cells. In practice, the effects are unpredictable, with varying levels of success. The predominant effects of anti-VEGF therapies are decreased vessel leakiness (hydraulic conductivity), decreased vessel diameters and pruning of the immature vessel network. It is thought that each of these can influence perfusion of the vessel network, inducing flow in regions that were previously sluggish or stagnant. Unfortunately, when anti-VEGF therapies affect vessel structure and function, the changes are dynamic and overlapping in time, and it has been difficult to identify a consistent and predictable normalization ``window'' during which perfusion and subsequent drug delivery is optimal. This is largely due to the non-linearity in the system, and the inability to distinguish the effects of decreased vessel leakiness from those due to network structural changes in clinical trials or animal studies. We have developed a mathematical model to calculate blood flow in complex tumor networks imaged by two-photon microscopy. The model incorporates the necessary and sufficient components for addressing the problem of normalization of tumor vasculature: i) lattice-Boltzmann calculations of the full flow field within the vasculature and within the tissue, ii) diffusion and convection of soluble species such as oxygen or drugs within vessels and the tissue domain, iii) distinct and spatially-resolved vessel hydraulic conductivities and permeabilities for each species, iv) erythrocyte particles advecting in the flow and delivering oxygen with real oxygen release kinetics, v) shear stress-mediated vascular remodeling. This model, guided by multi-parameter intravital imaging of tumor vessel structure