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Sample records for degraded core coolability

  1. Coolability of severely degraded CANDU cores

    International Nuclear Information System (INIS)

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

    1995-07-01

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

  2. Coolability of severely degraded CANDU cores. Revised

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  3. LACOMERA - large scale experiments on core degradation, melt retention and coolability at the Forschungszentrum Karslruhe

    International Nuclear Information System (INIS)

    Miassoedov, A.; Alsmeyer, H.; Meyer, L.

    2003-01-01

    The LACOMERA project at the Forschungszentrum Karlsruhe is a 3 year shared-cost action within the Fifth Framework Programme which started in September 2002. The overall objectives of the LACOMERA project are to provide research institutions from the EU member countries and associated states access to large scale experimental facilities at the Forschungszentrum Karlsruhe which shall be used to increase the knowledge of the quenching of a degraded core and regaining melt coolability in the reactor pressure vessel, of possible melt dispersion to the cavity, of molten core concrete interaction and of ex-vessel melt coolability. One major aspect is to understand how these events affect the safety of European reactors so as to lead to soundly-based accident management procedures. The project will bring together interested partners of different European member states in the area of severe accident analysis and control, with the goal to increase the public confidence in the use of nuclear energy. Moreover, partners from the newly associated states should be included as far as possible, and therefore the needs of Eastern, as well as Western, reactors will be considered in LACOMERA project. The project offers a unique opportunity to get involved in the networks and activities supporting VVER safety, and for Eastern experts to get an access to large scale experimental facilities in a Western research organisation to improve understanding of material properties and core behaviour under severe accident conditions. As a result of the first call for proposals a project on air ingress test in the QUENCH facility has been selected. A second call for proposals is opened with a deadline of 31 December 2003. (author)

  4. Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

    1995-09-01

    Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

  5. Recent progress in the LACOMERA Project (Large-Scale Experiments on Core Degradation, Melt Retention and Coolability) at the Forschungszentrum Karslruhe

    International Nuclear Information System (INIS)

    Miassoedov, A.; Alsmeyer, H.; Eppinger, B.; Meyer, L.; Steinbrueck, M.

    2004-01-01

    The LACOMERA Project at the Forschungszentrum Karlsruhe (FZK) is a 3 year action within the 5 th Framework Programme of the EU. The overall objective of the project is to offer research institutions from the EU member countries and associated states access to four large-scale experimental facilities QUENCH, LIVE, DISCO-H, and COMET which can be used to investigate core melt scenarios from the beginning of core degradation to melt formation and relocation in the vessel, possible melt dispersion to the reactor cavity, and finally corium concrete interaction and corium coolability in the reactor cavity. As a result of two calls for proposals, seven organisations from four countries are expected to profit from the LACOMERA Project participating in preparation, conduct and analysis of the following experiments: QUENCH-L1: Air ingression impact on core degradation. The test has provided unique data for the investigation of air ingress phenomenology in conditions as representative as possible of the reactor case regarding the source term. QUENCH-L2: Boil-off of a flooded bundle. The test will be of a generic interest for all reactor types, providing a link between the severe accident and design basis areas, and would deliver oxidation and thermal hydraulic data at high temperatures. LIVE-L1: Simulation of melt relocation into the Reactor Pressure Vessel (RPV) lower head for VVER conditions. The experiment will provide important information on the melt pool behaviour during the stages of air circulation at the outer RPV surface with a subsequent flooding of the lower head. LIVE-L2: Transient corium spreading and its impact on the heat fluxes to the RPV wall and on the final shape of the melt in the RPV lower head. The test will address the questions of melt stabilisation and the effects of crust formation near the RPV wall for a nonsymmetrical melt pool shape. COMET-L1: Long-term 2D concrete ablation in siliceous concrete cavity at intermediate decay heat power level with

  6. Post-accident core coolability of light water reactors

    International Nuclear Information System (INIS)

    Michio, I.; Teruo, I.; Tomio, Y.; Tsutao, H.

    1983-01-01

    A study on post-accident core coolability of LWR is discussed based on the practical fuel failure behavior experienced in NSRR, PBF, PNS and others. The fuel failure behavior at LOCA, RIA and PCM conditions are reviewed, and seven types of fuel failure modes are extracted as the basic failure mechanism at accident conditions. These are: cladding melt or brittle failure, molten UO 2 failure, high temperature cladding burst, low temperature cladding burst, failure due to swelling of molten UO 2 , failure due to cracks of embrittled cladding for irradiated fuel rods, and TMI-2 core failure. The post-accident core coolability at each failure mode is discussed. The fuel failures caused actual flow blockage problems. A characteristic which is common among these types is that the fuel rods are in the conditions violating the present safety criteria for accidents, and UO 2 pellets are in melting or near melting hot conditions when the fuel rods failed

  7. 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

  8. Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

    International Nuclear Information System (INIS)

    Jasiulevicius, Audrius

    2003-01-01

    This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

  9. Proceedings of the Workshop on in-vessel core debris retention and coolability

    International Nuclear Information System (INIS)

    1999-01-01

    This conference on in-vessel core debris retention and coolability is composed of 37 papers grouped in three sessions: session 1 (Keynote papers: Key phenomena of late phase core melt progression, accident management strategies and status quo of severe fuel damage codes, In-vessel retention as a severe accident management scheme, GAREC analyses in support of in-vessel retention concept, Latest findings of RASPLAV project); session 2 - Experiments and model development with five sub-sessions: sub-session 1 (Debris bed heat transfer: Debris and Pool Formation/Heat Transfer in FARO-LWR: Experiments and Analyses, Evaporation and Flow of Coolant at the Bottom of a Particle-Bed modelling Relocated Debris, Investigations on the Coolability of Debris in the Lower Head with WABE-2D and MESOCO-2D, Uncertainty and Sensitivity Analysis of the Heat Transfer Mechanisms in the Lower Head, Simulation of the Arrival and Evolution of Debris in a PWR Lower Head with the SFD ICARE2 code), sub-session 2 (Corium properties, molten pool natural convection, and crust formation: Physico-chemistry and corium properties for in-vessel retention, Experimental data on heat flux distribution from volumetrically heated pool with frozen boundaries, Thermal hydraulic phenomena in corium pools - numerical simulation with TOLBIAC and experimental validation with BALI, TOLBIAC code simulations of some molten salt RASPLAV experiments, SIMECO experiments on in-vessel melt pool formation and heat transfer with and without a metallic layer, Numerical investigation of turbulent natural convection heat transfer in an internally-heated melt pool and metallic layer, Current status and validation of CON2D and 3D code, Free convection of heat-generating fluid in a constrained during experimental simulation of heat transfer in slice geometry), sub-session 3 (Gap formation and gap cooling: Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water, Experimental investigations

  10. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  11. State-of-the-Art Report on Molten Corium Concrete Interaction and Ex-Vessel Molten Core Coolability

    International Nuclear Information System (INIS)

    Bonnet, Jean-Michel; Cranga, Michel; Vola, Didier; Marchetto, Cathy; Kissane, Martin; ); Robledo, Fernando; Farmer, Mitchel T.; Spengler, Claus; Basu, Sudhamay; Atkhen, Kresna; Fargette, Andre; Fisher, Manfred; Foit, Jerzi; Hotta, Akitoshi; Morita, Akinobu; Journeau, Christophe; Moiseenko, Evgeny; Polidoro, Franco; Zhou, Quan

    2017-01-01

    Activities carried out over the last three decades in relation to core-concrete interactions and melt coolability, as well as related containment failure modes, have significantly increased the level of understanding in this area. In a severe accident with little or no cooling of the reactor core, the residual decay heat in the fuel can cause the core materials to melt. One of the challenges in such cases is to determine the consequences of molten core materials causing a failure of the reactor pressure vessel. Molten corium will interact, for example, with structural concrete below the vessel. The reaction between corium and concrete, commonly referred to as MCCI (molten core concrete interaction), can be extensive and can release combustible gases. The cooling behaviour of ex-vessel melts through sprays or flooding is also complex. This report summarises the current state of the art on MCCI and melt coolability, and thus should be useful to specialists seeking to predict the consequences of severe accidents, to model developers for severe-accident computer codes and to designers of mitigation measures

  12. Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Jasiulevicius, Audrius

    2003-07-01

    This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

  13. PWR degraded core analysis

    International Nuclear Information System (INIS)

    Gittus, J.H.

    1982-04-01

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

  14. The results of the CCI-3 reactor material experiment investigating 2-D core-concrete interaction and debris coolability with a siliceous concrete crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs for future plants. Despite years of international research, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a summary description of the test facility and an overview of test results

  15. Experimental study on coolability of particulate core-metal debris bed with oxidization, (2). Fragmentation and enhanced heat transfer in zircaloy debris bed

    International Nuclear Information System (INIS)

    Su, Guanghui; Sugiyama, Ken-ichiro; Aoki, Hiroomi; Kimura, Iichi

    2006-01-01

    The oxidization and coolability characteristics of the particulate Zircaloy debris bed, which is deposited under the hard debris and through which first vapor penetrates and then water penetrates, are studied in the present paper. In the vapor penetration experiments, it is found that Zircaloy debris particles are effectively broken into small pieces after making thick oxidized layer with deep clacks by rapid oxidization under the condition that vapor with 20 cm/s penetrates for 30 to 70 min at an initial debris bed temperature of 1,030degC. It is also confirmed in the water penetration experiments that the oxidized particle debris bed has potentially of high coolability when water penetrates through the fully oxidized particle bed because of a high capillary force originating from those particles with deep cracks on their surfaces. Based on the present study, a new scenario for the appearance and disappearance of the hot spot in the TMI-2 accident is possible. The particulate core-metal core-metal debris bed is first heated up by rapid oxidization with heat generation when vapor can penetrate through the debris bed with porosities. This corresponds to the appearance of the hot spot. The resultant oxidized particulate debris bed causes a high coolability due to its high capillary force when the water can touch the debris bed at wet condition. This corresponds to the disappearance of the hot spot. (author)

  16. The Results of the CCI-3 Reactor Material Experiment Investigating 2-D Core-Concrete Interaction and Debris Coolability with a Siliceous Concrete Crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Lomperski, S.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program conducted reactor materials experiments and associated analysis to achieve the following two objectives: 1) resolve the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs of future plants. With respect to the second objective, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion profiles predicted by codes such as WECHSL, COSACO, TOLBIAC, MEDICIS, and MELCOR. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test investigated the long-term interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined time interval, the cavity was flooded with water to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the facility and an overview of results from this test. (authors)

  17. Coolability in the frame of core melt accidents in light water reactors. Model development and validation for ATHLET-CD and ASTEC. Final report; Kuehlbarkeit im Rahmen von Kernschmelzunfaellen bei Leichtwasserreaktoren. Modellentwicklung und Validierung fuer ATHLET-CD und ASTEC. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Michael; Pohlner, Georg; Rahman, Saidur; Berkhan, Ana

    2015-07-15

    The code system ATHLET/ATHLET-CD is being developed in the frame of the reactor safety research of the German Federal Ministry for Economic Affairs and Energy (BMWi) within the topic analysis of transients and accident sequences. It serves for simulation of transients and accidents to be used in safety analyses for light water reactors. In the present project the development and validation of models for ATHLET-CD for description of the processes during severe accidents are continued. These works should enable broad safety analyses by a mechanistic description of the processes even during late phases of a degrading core and by this a profound estimation on coolability and accident management options during every phase. With the actual status of modelling in ATHLET-CD analyses on coolability are made to give a solid base for estimates about stabilization by cooling or accident progression, dependent on the scenario. The modeling in the MEWA module, describing the processes in a severely degraded core in ATHLET-CD, is extended on the processes in the lower plenum. For this, the model on melt pool behavior is extended and linked to the RPV wall. The coupling between MEWA and the thermal-hydraulics of ATHLET-CD is improved. The validation of the models is continued by calculations on new experiments and comparing analyses done in the frame of the European Network SARNET-2. For the European integral code ASTEC contributions from the modeling for ATHLET-CD will be done, especially by providing a model for the melt behavior in the lower plenum of a LWR. This report illustrates the work carried out in the frame of this project, and shows results of calculations and the status of validation by recalculations on experiments for debris bed coolability, melt pool behavior as well as jet fragmentation and debris bed formation.

  18. Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR

    Energy Technology Data Exchange (ETDEWEB)

    Galushin, Sergey, E-mail: galushin@kth.se; Kudinov, Pavel, E-mail: pkudinov@kth.se

    2016-12-15

    Highlights: • A data base of the debris properties in lower plenum generated using MELCOR code. • The timing of safety systems has significant effect on the relocated debris properties. • Loose coupling between core relocation and vessel failure analyses was established. - Abstract: Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario

  19. 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

  20. Core degradation and fission product release

    International Nuclear Information System (INIS)

    Wright, R.W.; Hagen, S.J.L.

    1992-01-01

    Experiments on core degradation and melt progression in severe LWR accidents have provided reasonable understanding of the principal processes involved in the early phase of melt progression that extends through core degradation and metallic material melting and relocation. A general but not a quantitative understanding of late phase melt progression that involves ceramic material melting and relocation has also been obtained, primarily from the TMI-2 core examination. A summary is given of the current state of knowledge on core degradation and melt progression obtained from these integral experiments and of the principal remaining significant uncertainties. A summary is also given of the principal results on in-vessel fission product release obtained from these experiments. (author). 8 refs, 5 figs, 3 tabs

  1. Degraded core studies at INEL

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  2. Effects of thermohydraulics on clad ballooning, flow blockage and coolability in a LOCA

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Neitzel, H.J.; Wiehr, K.

    1983-01-01

    Thermohydraulic boundary conditions have a dominating effect on clad ballooning, flow blockage and coolability: Increasing heat transfer to the fluid decreases the total circumferential strain; Countercurrent flow in a combined injection leads to a relatively small flow blockage; Burst claddings exhibit premature quenching. Differences in the test results obtained in several countries are mainly due to different thermohydraulic test conditions; all test data are consistent with the understanding elaborated within the REBEKA program. Core coolability in a LOCA can be maintained. (author)

  3. Managing water addition to a degraded core

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  4. Coolability of volumetrically heated particle beds

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Muhammad

    2017-03-22

    In case of a severe nuclear reactor accident, with loss of coolant, a particle bed may be formed from the fragmentation of the molten core in the residual water at different stages of the accident. To avoid further propagation of the accident and maintain the integrity of the reactor pressure vessel, the decay heat of the particle bed must be removed. To better understand the various thermo-hydraulic processes within such heat-generating particle beds, the existing DEBRIS test facility at IKE has been modified to be able to perform novel boiling, dryout and quenching experiments. The essential experimental data includes the pressure gradients measured by 8 differential pressure transducers along the bed height as a function of liquid and vapour superficial velocities, the determination of local dryout heat fluxes for different system pressures as well as the local temperature distribution measured by a set of 51 thermocouples installed inside the particle bed. The experiments were carried out for two different particle beds: a polydispersed particle bed which consisted of stainless steel balls (2 mm, 3 mm and 6 mm diameters) and an irregular particle bed which consisted of a mixture of steel balls (3 mm and 6 mm) and irregularly shaped Al{sub 2}O{sub 3} particles. Additionally, all experiments were carried out for different flow conditions, such as the reference case of passive 1D top-flooding, 1D bottom flooding (driven by external pumps and different downcomer configurations) and 2D top-/bottom-/lateral flooding with a perforated downcomer. In this work, it has been observed that for both particle beds with downcomer configurations an open downcomer leads to the best coolability (dryout heat flux = 1560 kW/m{sup 2}, polydispersed particle bed, psys = 1 bar) of the particle bed, mainly due to bottom-flow with enhanced natural convection. It has also been shown that a potential lateral flow via a perforation of the downcomer does not bring any further improvements

  5. Coolability of volumetrically heated particle beds

    International Nuclear Information System (INIS)

    Rashid, Muhammad

    2017-01-01

    In case of a severe nuclear reactor accident, with loss of coolant, a particle bed may be formed from the fragmentation of the molten core in the residual water at different stages of the accident. To avoid further propagation of the accident and maintain the integrity of the reactor pressure vessel, the decay heat of the particle bed must be removed. To better understand the various thermo-hydraulic processes within such heat-generating particle beds, the existing DEBRIS test facility at IKE has been modified to be able to perform novel boiling, dryout and quenching experiments. The essential experimental data includes the pressure gradients measured by 8 differential pressure transducers along the bed height as a function of liquid and vapour superficial velocities, the determination of local dryout heat fluxes for different system pressures as well as the local temperature distribution measured by a set of 51 thermocouples installed inside the particle bed. The experiments were carried out for two different particle beds: a polydispersed particle bed which consisted of stainless steel balls (2 mm, 3 mm and 6 mm diameters) and an irregular particle bed which consisted of a mixture of steel balls (3 mm and 6 mm) and irregularly shaped Al 2 O 3 particles. Additionally, all experiments were carried out for different flow conditions, such as the reference case of passive 1D top-flooding, 1D bottom flooding (driven by external pumps and different downcomer configurations) and 2D top-/bottom-/lateral flooding with a perforated downcomer. In this work, it has been observed that for both particle beds with downcomer configurations an open downcomer leads to the best coolability (dryout heat flux = 1560 kW/m 2 , polydispersed particle bed, psys = 1 bar) of the particle bed, mainly due to bottom-flow with enhanced natural convection. It has also been shown that a potential lateral flow via a perforation of the downcomer does not bring any further improvements in

  6. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  7. On the air coolability of TRIGA reactors following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

    1986-01-01

    This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

  8. Investigation of debris bed formation, spreading and coolability

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, P.; Konovalenko, A.; Grishchenko, D.; Yakush, S.; Basso, S.; Lubchenko, N.; Karbojian, A. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    The work is motivated by the severe accident management strategy adopted in Nordic type BWRs. It is assumed that core melt ejected from the vessel will fragment, quench and form a coolable debris bed in a deep water pool below the vessel. In this work we consider phenomena relevant to the debris bed formation and coolability. Several DEFOR-A (Debris Bed Formation - Agglomeration) tests have been carried out with new corium melt material and a melt releasing nozzle mockup. The influence of the melt material, melt superheat, jet free fall height on the (i) faction of agglomerated debris, (ii) particle size distribution, (iii) ablation/plugging of the nozzle mockup has been addressed. Results of the DECOSIM (Debris Coolability Simulator) code validation against available COOLOCE data are presented in the report. The dependence of DHF on system pressure from COOLOCE experiments can be reproduced quite accurately if either the effective particle diameter or debris bed porosity is increased. For a cylindrical debris bed, good agreement is achieved in DECOSIM simulations for the particle diameter 0.89 mm and porosity 0.4. The results obtained are consistent with MEWA simulation where larger particle diameters and porosities were found to be necessary to reproduce the experimental data on DHF. It is instructive to note that results of DHF prediction are in better agreement with POMECO-HT data obtained for the same particles. It is concluded that further clarification of the discrepancies between different experiments and model predictions. In total 13 exploratory tests were carried out in PDS (particulate debris spreading) facility to clarify potential influence of the COOLOCE (VTT) facility heaters and TCs on particle self-leveling process. Results of the preliminary analysis suggest that there is no significant influence of the pins on self-leveling, at least for the air superficial velocities ranging from 0.17 up to 0.52 m/s. Further confirmatory tests might be needed

  9. Investigation of debris bed formation, spreading and coolability

    International Nuclear Information System (INIS)

    Kudinov, P.; Konovalenko, A.; Grishchenko, D.; Yakush, S.; Basso, S.; Lubchenko, N.; Karbojian, A.

    2013-08-01

    The work is motivated by the severe accident management strategy adopted in Nordic type BWRs. It is assumed that core melt ejected from the vessel will fragment, quench and form a coolable debris bed in a deep water pool below the vessel. In this work we consider phenomena relevant to the debris bed formation and coolability. Several DEFOR-A (Debris Bed Formation - Agglomeration) tests have been carried out with new corium melt material and a melt releasing nozzle mockup. The influence of the melt material, melt superheat, jet free fall height on the (i) faction of agglomerated debris, (ii) particle size distribution, (iii) ablation/plugging of the nozzle mockup has been addressed. Results of the DECOSIM (Debris Coolability Simulator) code validation against available COOLOCE data are presented in the report. The dependence of DHF on system pressure from COOLOCE experiments can be reproduced quite accurately if either the effective particle diameter or debris bed porosity is increased. For a cylindrical debris bed, good agreement is achieved in DECOSIM simulations for the particle diameter 0.89 mm and porosity 0.4. The results obtained are consistent with MEWA simulation where larger particle diameters and porosities were found to be necessary to reproduce the experimental data on DHF. It is instructive to note that results of DHF prediction are in better agreement with POMECO-HT data obtained for the same particles. It is concluded that further clarification of the discrepancies between different experiments and model predictions. In total 13 exploratory tests were carried out in PDS (particulate debris spreading) facility to clarify potential influence of the COOLOCE (VTT) facility heaters and TCs on particle self-leveling process. Results of the preliminary analysis suggest that there is no significant influence of the pins on self-leveling, at least for the air superficial velocities ranging from 0.17 up to 0.52 m/s. Further confirmatory tests might be needed

  10. Improvement of core degradation model in ISAAC

    International Nuclear Information System (INIS)

    Kim, Dong Ha; Kim, See Darl; Park, Soo Yong

    2004-02-01

    If water inventory in the fuel channels depletes and fuel rods are exposed to steam after uncover in the pressure tube, the decay heat generated from fuel rods is transferred to the pressure tube and to the calandria tube by radiation, and finally to the moderator in the calandria tank by conduction. During this process, the cladding will be heated first and ballooned when the fuel gap internal pressure exceeds the primary system pressure. The pressure tube will be also ballooned and will touch the calandria tube, increasing heat transfer rate to the moderator. Although these situation is not desirable, the fuel channel is expected to maintain its integrity as long as the calandria tube is submerged in the moderator, because the decay heat could be removed to the moderator through radiation and conduction. Therefore, loss of coolant and moderator inside and outside the channel may cause severe core damage including horizontal fuel channel sagging and finally loss of channel integrity. The sagged channels contact with the channels located below and lose their heat transfer area to the moderator. As the accident goes further, the disintegrated fuel channels will be heated up and relocated onto the bottom of the calandria tank. If the temperature of these relocated materials is high enough to attack the calandria tank, the calandria tank would fail and molten material would contact with the calandria vault water. Steam explosion and/or rapid steam generation from this interaction may threaten containment integrity. Though a detailed model is required to simulate the severe accident at CANDU plants, complexity of phenomena itself and inner structures as well as lack of experimental data forces to choose a simple but reasonable model as the first step. ISAAC 1.0 was developed to model the basic physicochemical phenomena during the severe accident progression. At present, ISAAC 2.0 is being developed for accident management guide development and strategy evaluation. In

  11. Coolability of oxidized particulate debris bed accumulated in horizontal narrow gaps

    International Nuclear Information System (INIS)

    Arai, Y.; Sugiyama, K.; Narabayashi, T.

    2007-01-01

    When LOCA occurs in a nuclear reactor system, the coolability of the core would be kept as reported at a series of presentations in ICONE14. Therefore the probability of the core meltdown is negligible small. However, from the view point of defense in depth, it is necessary to be sure that the coolability of the bottom of reactor pressure vessel (RPV) is maintained even if a part of the core should melt and a substantial amount of debris should be deposited on the lower plenum. We carried out an experimental study in order to observe the coolability of particulate core-metal debris bed with 12 mm thickness accompanied with rapid heat generation because of oxidization, which was reported at ICONE14. The coolability was assured by a small amount of coolant supply because of high capillary force of oxidized fine particulate debris produced. In the present study, we examined the coolability of particulate debris bed deposited in narrower gap of 1 mm or 5 mm that coolant supply is hard. The particulate debris beds were piled up on the stainless steel sheet with 0.1 mm thickness, which was used to measure the bottom temperatures of particulate debris bed by using a thermo-video camera. We set up a heat supply section with heat input of 2.1 kW, which simulates the hard debris bed deposited on the particulate debris bed as reported for the TMI-2 accident. We measured the temperatures of the bottom surface of the heat supply section and the heat fluxes released into debris bed as well as the temperatures at the bottom of debris bed on the stainless steel sheet. It is found that when only the upper surface of particulate debris bed is in the film boiling, capillary force causes coolant supply to the particulate debris bed. Therefore, in the condition of thicker gap with small particulate debris, coolability of debris bed is improved. We find out that smaller particulate debris is moved by vapor movement. As a result, the area that high capillary force is caused because of

  12. Ex-vessel corium coolability sensitivity study with the CORQUENCH code

    International Nuclear Information System (INIS)

    Robb, Kevin; Corradini, Michael

    2009-01-01

    An unresolved safety issue for light water reactor beyond design basis accidents is the coolability and stabilization of ex-vessel core melt debris by top flooding. Several experimental programs, including the OECD MACE, MCCI-1, and the current MCCI-2 program, have investigated core-concrete interactions and debris cooling of ex-vessel core melts. As part of the OECD programs, the CORQUENCH computer model was developed based on phenomena identified from the experiments. Predictions by CORQUENCH have previously been compared against experiments and have also been extrapolated to reactor scale. The current study applied statistical techniques to investigate the importance of initial system parameters and cooling phenomena in CORQUENCH 3.01 on the accident progression of ex-vessel core melts. The purpose of this sensitivity study is to identify parameters that are of major importance, any code peculiarities over the range of inputs, and where modeling improvements may produce the most gain in prediction accuracy. The sensitivity studies were carried out over a range of input conditions, in 1-D and 2-D geometries, and for two concrete compositions. In terms of initial system parameters, the melt height had the most importance on concrete ablation and melt coolability. With respect to cooling phenomena, the amount of melt entrainment through the crust had the most importance on concrete ablation and melt coolability. (author)

  13. Study on coolability of melt pool with different strategies

    International Nuclear Information System (INIS)

    Kulkarni, P.P.; Nayak, A.K.

    2014-01-01

    Highlights: • Experiments have been performed to test quenching of molten pool with different schemes. • Top flooding, bottom flooding and indirect cooling schemes were used. • A single simulant material with same mass and initial temperature was used. • Bottom flooding technique is found to be the most effective technique. • A comparison of all the three techniques has been presented. - Abstract: After the Fukushima accident, there have been a lot of concerns regarding long term core melt stabilization following a severe accident in nuclear reactors. Several strategies have been contemplated for quenching and stabilization of core melt like top flooding, bottom flooding, indirect cooling, etc. However, the effectiveness of these schemes is yet to be determined properly, for which, lot of experiments are needed. Several experiments have been performed for coolability of molten pool under top flooding condition. A few experiments have been performed for study of coolability of melt pool under bottom flooding as well as for indirect cooling. Besides, these tests are very scattered because they involve different simulant materials, initial temperatures and masses of melt, which makes it very difficult to judge the effectiveness of a particular technique and advantage over the other. In the present paper we have carried out different experiments wherein a single simulant material with same mass was cooled with different techniques starting from the same initial temperature. The result showed that, while top flooding and indirect cooling took several hours to cool, bottom flooding took a few minutes to cool the melt which makes it the most effective technique

  14. Assessment of the integral code ASTEC with respect to the late in-vessel phase of core degradation

    International Nuclear Information System (INIS)

    D'Alessandro, Christophe; Starflinger, Joerg

    2014-01-01

    The integral code ASTEC is being developed jointly by GRS and IRSN as the European reference code for severe accidents. In the EU project CESAM it is foreseen to assess the capabilities of ASTEC to deal with a broad range of reactor designs (PWR, BWR, VVER, CANDU, Gen III+, etc.) and especially to model and capture the effect of severe accident mitigation measures. This requires a physically sound and sufficiently accurate modelling of the processes and phenomena that govern the course of the accident, and the modelling has to be validated to a sufficient extent. Concentrating on the in-vessel aspects of severe accidents, the present paper addresses these requirements by presenting results of ASTEC calculations for relevant experiments that cover the major physical phenomena during core degradation (melting and relocation of the fuel, oxidation, molten corium pool formation and its coolability in the lower plenum once it slumped from the core region). The assessment of models for bundle degradation is based on CORA (13 and W2). CORA represented a bundle of non-irradiated, electrically heated UO 2 -rods. Melt progression in strongly degraded geometry is addressed in the PHEBUS-FTP4 experiment carried out with irradiated fuel in debris bed configuration. The validation of molten pool modelling is based on BALI and RASPLAV-Salt experiments. The BALI-facility consists of a full-scale slice of lower plenum (allowing experiments at prototypical Rayleigh numbers) and utilizes uniformly heated water as simulant for corium. The RASPLAV experiments use a scaled-down slice of the lower head. Use of non-eutectic molten salt as simulant should address the effect of a significant solidification range typical for real corium. Calculation results of ASTEC are discussed in comparison with experimental measurements. Further, questions concerning the extrapolation of findings from validation to reactor application are critically discussed, concerning e.g. choice of model parameters

  15. Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

    International Nuclear Information System (INIS)

    Cho, Dae H.; Page, Richard J.; Abdulla, Sherif H.; Anderson, Mark H.; Klockow, Helge B.; 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 goal of our work is to provide the fundamental understanding 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. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb-Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed

  16. Technical Issues Associated with Air Ingression During Core Degradation

    International Nuclear Information System (INIS)

    Powers, Dana A.

    2000-01-01

    This paper has shown that it is possible to get significant air intrusion into a ruptured reactor vessel even from a reactor cavity with restricted access. This suggests that there is some importance to considering the consequences of air intrusion following vessel penetration by core debris. The consequences will depend on the nature of core degradation in air and other oxidizing gases. If, indeed, fuel becomes exposed to strongly oxidizing gases, significant releases of ruthenium and hexavalent urania can be expected. Hexavalent urania could alter the nature of cesium release and cesium revaporization from the reactor coolant system. Hexavalent urania could destabilize CSI and enhance the formation of gaseous iodine unless there are other materials that will react readily with atomic iodine along the flow path to the reactor containment

  17. In-vessel core degradation code validation matrix

    International Nuclear Information System (INIS)

    Haste, T.J.; Adroguer, B.; Gauntt, R.O.; Martinez, J.A.; Ott, L.J.; Sugimoto, J.; Trambauer, K.

    1996-01-01

    The objective of the current Validation Matrix is to define a basic set of experiments, for which comparison of the measured and calculated parameters forms a basis for establishing the accuracy of test predictions, covering the full range of in-vessel core degradation phenomena expected in light water reactor severe accident transients. The scope of the review covers PWR and BWR designs of Western origin: the coverage of phenomena extends from the initial heat-up through to the introduction of melt into the lower plenum. Concerning fission product behaviour, the effect of core degradation on fission product release is considered. The report provides brief overviews of the main LWR severe accident sequences and of the dominant phenomena involved. The experimental database is summarised. These data are cross-referenced against a condensed set of the phenomena and test condition headings presented earlier, judging the results against a set of selection criteria and identifying key tests of particular value. The main conclusions and recommendations are listed. (K.A.)

  18. Coolability of the melt in the reactor tank. A compilation and evaluation of the state of the art and suggestions for experiments in the area; Smaeltans kylbarhet i reaktortanken. En sammanstaellning och vaerdering av kunskapslaeget och foerslag till experiment inom omraadet

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Ferenc [ES-konsult Energi och Saekerhet AB, Stockholm (Sweden)

    2002-04-01

    This study is limited to experiments about phenomena and mechanisms that affect the coolability of core debris in the reactor tank that may delay the tank rupture. The goal of the study is to get an overview of the phenomena that are important for the in-vessel coolability, and to evaluate the need for new experiments. Both theoretical and experimental projects are suggested.

  19. Experimental investigation of multidimensional cooling effects on the coolability of a debris bed

    International Nuclear Information System (INIS)

    Rashidi, M.; Kulenovici, R.; Laurieni, E.

    2011-01-01

    During a severe accident in a light water reactor, the core can melt and be relocated to the lower plenum of the reactor pressure vessel. There it can form a particulate debris bed due to the possible presence of water. Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long-term coolability of debris beds, the down-scaled non nuclear test facility DEBRIS has been established at IKE. The major objectives of the experimental investigations at this test facility are the determination of local pressure drops for steady state boiling to check friction laws, the determination of dryout heat fluxes under various conditions for validation of numerical models, and the analysis of quenching processes of dry hot debris beds. A large number of 1D-experiments were carried out to investigate the coolability limits for different bed configurations at various thermohydraulic conditions, and to validate numerical models which can be used in reactor safety studies. Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favored. This paper presents 2D experimental results, based on various kinds of water inflow conditions into the bed, boiling and dryout tests with different bed configurations and different system pressures. Preliminary results show that the system pressure has no significant effect on the fundamental shape of the pressure gradient inside the bed, whereas with increasing system pressure the coolability limits are increased

  20. VVER-specific features regarding core degradation - Status Report

    International Nuclear Information System (INIS)

    Hozer, Z.; Trambauer, K.; Duspiva, J.

    1999-01-01

    The objective of this report is to compare VVER reactors to pressurised water reactors (PWRs) of Western design from the point of view of core degradation phenomena using the terminology which was applied to the systematisation of severe accident phenomena in earlier CSNI reports. In the following the acronym 'PWR' is used for a PWR of Western design. The basic design features are described and the most important parameters are summarised in order to identify the differences between the two reactor types. In some specific cases the comparison shows more similarities with boiling water reactors (BWRs) than with PWRs. The known VVER experimental support is also summarised. RBMKs are not included in this report, as this reactor type is not operated in OECD countries, furthermore its design is completely different from those of VVERs and PWRs. The scope of this report is limited to in-vessel severe fuel damage phenomena. Neither thermal hydraulic processes involving no core degradation, nor containment phenomena, are discussed in detail. The VVER (water-cooled water-moderated power reactor) is a pressurised light water reactor of Soviet design. It operates on the same principles as a Western PWR reactor and uses similar technological systems. The primary coolant is pressurised water, which heats up in the reactor core and steam is produced on the secondary side of steam generators. The comparison of basic geometrical and technological parameters pointed out some differences between a PWR and a VVER, but it should be noted that differences exist even between two Western PWRs of different design. The VVER reactors are special types of PWRs, the most important design features of which are the horizontal steam generators and the hexagonal core structure. Similarity between PWR and VVER reactors was found in the comparison of dominant accidents sequences leading to core melt. The accident progression sequence consists of the same steps for VVERs and PWRs. The larger water

  1. Debris bed coolability using a 3-D two phase model in a porous medium

    Energy Technology Data Exchange (ETDEWEB)

    Bechaud, C.; Duval, F.; Fichot, F. [CEA Cadarache, Inst. de Protection et de Surete Nucleaire13 - Saint-Paul-lez-Durance (France); Quintard, M. [Institut de Mecanique des Fluides de Toulouse, 31 (France); Parent, M. [CEA Grenoble, Dept. de Thermohydraulique et de Physique, 38 (France)

    2001-07-01

    During a severe nuclear accident, a part of the molten corium resulting from the core degradation may relocate in the lower plenum of the reactor vessel. In order to predict the safety margin of the reactor under such conditions, the coolability of this porous heat-generating medium is evaluated in this study and compared with other investigations. In this work, conservation equations derived for debris beds are implemented in the three dimensional thermal-hydraulic module of the CATHARE code. The coolant flow is a two phase flow with phase change. The momentum balance equation for each fluid phase is an extension of Darcy's law. This extension takes into account the capillary effects between the two phases, the relative permeabilities and passabilities of each phase, the interfacial drag force between liquid and gas, and the porous bed configuration (porosity, particle diameter,... ). The model developed is three-dimensional which is important to better predict the flow in configuration such as counter-current flow or to emphasize preferential ways induced by porous geometry. The energy balance equations of the three phases (liquid, gas and solid phase) are obtained by a volume averaging process of the local conservation equations. In this method, the local thermal non-equilibrium between the three phases is considered and the heat exchanges, the phase change rate as well as the thermal dispersion coefficients are calculated as a function of the local geometry of the porous medium. Such a method allows the numerical estimation of these thermal properties which are very difficult to determine experimentally. This feature is a great advantage of this approach. After a brief description of the thermal-hydraulic model, one-dimensional predictions of critical dryout fluxes are presented and compared with results from the literature. Reasonable agreement is obtained. Then a two-dimensional calculation is presented and shows the influence of the porous medium

  2. Fuel and control rod failure behavior during degraded core accidents

    International Nuclear Information System (INIS)

    Chung, K.S.

    1984-01-01

    As a part of the pretest and posttest analyses of Light Water Reactor Source Term Experiments (STEP) which are conducted in the Transient Reactor Test (TREAT) facility, this paper investigates the thermodynamic and material behaviors of nuclear fuel pins and control rods during severe core degradation accidents. A series of four STEP tests are being performed to simulate the characteristics of the power reactor accidents and investigate the behavior of fission product release during these accidents. To determine the release rate of the fission products from the fuel pins and the control rod materials, information concerning the timing of the clad failure and the thermodynamic conditions of the fuel pins and control rods are needed to be evaluated. Because the phase change involves a large latent heat and volume expansion, and the phase change is a direct cause of the clad failure, the understanding of the phase change phenomena, particularly information regarding how much of the fuel pin and control rod materials are melted are very important. A simple energy balance model is developed to calculate the temperature profile and melt front in various heat transfer media considering the effects of natural convection phenomena on the melting and freezing front behavior

  3. PIV Visualization of Bubble Induced Flow Circulation in 2-D Rectangular Pool for Ex-Vessel Debris Bed Coolability

    Energy Technology Data Exchange (ETDEWEB)

    Han, Teayang; Kim, Eunho; Park, Hyun Sun; Moriyama, Kiyofumi [POSTECH, Pohang (Korea, Republic of)

    2015-10-15

    The previous research works demonstrated the debris bed formation on the flooded cavity floor in experiments. Even in the cases the core melt is once solidified, the debris bed can be re-melted due to the decay heat. If the debris bed is not cooled enough by the coolant, the re-melted debris bed will react with the concrete base mat. This situation is called the molten core-concrete interaction (MCCI) which threatens the integrity of the containment by generated gases which pressurize the containment. Therefore securing the long term coolability of the debris bed in the cavity is crucial. According to the previous research works, the natural convection driven by the rising bubbles affects the coolability and the formation of the debris bed. Therefore, clarification of the natural convection characteristics in and around the debris bed is important for evaluation of the coolability of the debris bed. In this study, two-phase flow around the debris bed in a 2D slice geometry is visualized by PIV method to obtain the velocity map of the flow. The DAVINCI-PIV was developed to investigate the flow around the debris bed. In order to simulate the boiling phenomena induced by the decay heat of the debris bed, the air was injected separately by the air chamber system which consists of the 14 air-flowmeters. The circulation flow developed by the rising bubbles was visualized by PIV method.

  4. Transients analysis able to lead Pressurised Water Reactors cores to degraded situations, analysis of resulting configurations

    International Nuclear Information System (INIS)

    Shin, Hyeong-Ki

    1999-01-01

    The severe accidents that occurred recently on nuclear reactors such as Chernobyl and T.M.1.2 have led many countries utilizing nuclear energy to examine their severe accident management. This thesis focuses on this problem and aims at analyzing, in terms of reactivity, degraded core behavior resulting from different accidental configurations. Two types of core degradation can be encountered: local degradation (the destruction of isolated assemblies in the core) or spreading degradation (the destruction of neighboring assemblies). The TMI accident is an example of spreading degradation in the core. The simplicity of implementing the control rod ejection accident calculation as compared to other accidental transients have motivated the choice of this accident as a determinant for local degraded core configurations. The control rod ejection accident presents important three dimensional effects and introduces neutronic/thermohydraulic coupling. The implementation and validation of already existing three dimensional coupled calculation scheme, allowed one to analyze the consequences of such an accident and to the conclusion that only unrealistic hypotheses of assembly permutation could lead to a partial core degradation. A reasonable estimate of stored energy in the assemblies with high bum up, in relation to the stored energy in the hot spot, was also obtained for the first time. The recently performed experiments (CABRI experiments) showed that in highly burned up assemblies, the capacity to store energy decreases strongly in relation to new assemblies. This first estimate of the distribution of produced energy between different assemblies, during the rod ejection accident, offers an important piece of knowledge in the study of the consequences of an eventual fuel cycle extension (presently under consideration by development companies). Finally, the analysis of degraded core reactivity itself has been performed for a vast range of the degraded core configurations

  5. In-vessel core degradation code validation matrix update 1996-1999. Report by an OECD/NEA group of experts

    International Nuclear Information System (INIS)

    2001-02-01

    In 1991 the Committee on the Safety of Nuclear Installations (CSNI) issued a State-of-the-Art Report (SOAR) on In-Vessel Core Degradation in Light Water Reactor (LWR) Severe Accidents. Based on the recommendations of this report a Validation Matrix for severe accident modelling codes was produced. Experiments performed up to the end of 1993 were considered for this validation matrix. To include recent experiments and to enlarge the scope, an update was formally inaugurated in January 1999 by the Task Group on Degraded Core Cooling, a sub-group of Principal Working Group 2 (PWG-2) on Coolant System Behaviour, and a selection of writing group members was commissioned. The present report documents the results of this study. The objective of the Validation Matrix is to define a basic set of experiments, for which comparison of the measured and calculated parameters forms a basis for establishing the accuracy of test predictions, covering the full range of in-vessel core degradation phenomena expected in light water reactor severe accident transients. The emphasis is on integral experiments, where interactions amongst key phenomena as well as the phenomena themselves are explored; however separate-effects experiments are also considered especially where these extend the parameter ranges to cover those expected in postulated LWR severe accident transients. As well as covering PWR and BWR designs of Western origin, the scope of the review has been extended to Eastern European (VVER) types. Similarly, the coverage of phenomena has been extended, starting as before from the initial heat-up but now proceeding through the in-core stage to include introduction of melt into the lower plenum and further to core coolability and retention to the lower plenum, with possible external cooling. Items of a purely thermal hydraulic nature involving no core degradation are excluded, having been covered in other validation matrix studies. Concerning fission product behaviour, the effect

  6. Status of degraded core issues. Synthesis paper prepared by G. Bandini in collaboration with the NEA task group on degraded core cooling

    International Nuclear Information System (INIS)

    2001-02-01

    The in-vessel evolution of a severe accident in a nuclear reactor is characterised, generally, by core uncover and heat-up, core material oxidation and melting, molten material relocation and debris behaviour in the lower plenum up to vessel failure. The in-vessel core melt progression involves a large number of physical and chemical phenomena that may depend on the severe accident sequence and the reactor type under consideration. Core melt progression has been studied in the last twenty years through many experimental works. Since then, computer codes are being developed and validated to analyse different reactor accident sequences. The experience gained from the TMI-2 accident also constitutes an important source of data. The understanding of core degradation process is necessary to evaluate initial conditions for subsequent phases of the accident (ex-vessel and within the containment), and define accident management strategies and mitigative actions for operating and advanced reactors. This synthesis paper, prepared within the Task Group on Degraded Core Cooling (TG-DCC) of PWG2, contains a brief summary of current views on the status of degraded core issues regarding light water reactors. The in-vessel fission product release and transport issue is not addressed in this paper. The areas with remaining uncertainties and the needs for further experimental investigation and model development have been identified. The early phase of core melt progression is reasonably well understood. Remaining uncertainties may be addressed on the basis of ongoing experimental activities, e.g. on core quenching, and research programs foreseen in the near future. The late phase of core melt progression is less understood. Ongoing research programs are providing additional valuable information on corium molten pool behaviour. Confirmatory research is still required. The pool crust behaviour and material relocation into the lower plenum are the areas where additional research should

  7. Physics of coolability of top flooded molten corium

    International Nuclear Information System (INIS)

    Kulkarni, P.P.; Singh, R.K.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    During a postulated severe accident in a nuclear reactor in case of ex-vessel scenario the molten corium can be relocated in the containment cavity forming a melt pool. In order to arrest further progression of severe accident, complete quenching of the molten corium pool is necessary. Most common way to deal with ex-vessel scenario is to flood the melt pool with large quantity of water. However, the mechanism of coolability is much more complex involving multi-component, multiphase heat, mass and momentum transfer. In this paper, a mechanistic model has been presented for the corium coolability under top flooding conditions. The model has been validated with the experimental data of COMECO test facility available in literature. Simulations have been carried out using the model to explore the physics behind the corium coolability with MCCI under top flooding condition. Variations in the thermo-physical properties as a result of MCCI have been considered and its effect on coolability has been studied. (author)

  8. 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

  9. Implications for accident management of adding water to a degrading reactor core

    International Nuclear Information System (INIS)

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J.

    1994-02-01

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents

  10. Implications for accident management of adding water to a degrading reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kuan, P.; Hanson, D.J.; Pafford, D.J.; Quick, K.S.; Witt, R.J. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-02-01

    This report evaluates both the positive and negative consequences of adding water to a degraded reactor core during a severe accident. The evaluation discusses the earliest possible stage at which an accident can be terminated and how plant personnel can best respond to undesired results. Specifically discussed are (a) the potential for plant personnel to add water for a range of severe accidents, (b) the time available for plant personnel to act, (c) possible plant responses to water added during the various stages of core degradation, (d) plant instrumentation available to understand the core condition and (e) the expected response of the instrumentation during the various stages of severe accidents.

  11. Stiffness and strength degradation of damaged truss core composites

    Czech Academy of Sciences Publication Activity Database

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

    2015-01-01

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

  12. Experimental investigation of the coolability of blocked hexagonal bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

    2017-06-15

    Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

  13. Acetylation-Mediated Proteasomal Degradation of Core Histones during DNA Repair and Spermatogenesis

    Science.gov (United States)

    Qian, Min-Xian; Pang, Ye; Liu, Cui Hua; Haratake, Kousuke; Du, Bo-Yu; Ji, Dan-Yang; Wang, Guang-Fei; Zhu, Qian-Qian; Song, Wei; Yu, Yadong; Zhang, Xiao-Xu; Huang, Hai-Tao; Miao, Shiying; Chen, Lian-Bin; Zhang, Zi-Hui; Liang, Ya-Nan; Liu, Shan; Cha, Hwangho; Yang, Dong; Zhai, Yonggong; Komatsu, Takuo; Tsuruta, Fuminori; Li, Haitao; Cao, Cheng; Li, Wei; Li, Guo-Hong; Cheng, Yifan; Chiba, Tomoki; Wang, Linfang; Goldberg, Alfred L.; Shen, Yan; Qiu, Xiao-Bo

    2013-01-01

    SUMMARY Histone acetylation plays critical roles in chromatin remodeling, DNA repair, and epigenetic regulation of gene expression, but the underlying mechanisms are unclear. Proteasomes usually catalyze ATP- and polyubiquitin-dependent proteolysis. Here we show that the proteasomes containing the activator PA200 catalyze the polyubiquitin-independent degradation of histones. Most proteasomes in mammalian testes (“spermatoproteasomes”) contain a spermatid/sperm-specific α-subunit α4s/PSMA8 and/or the catalytic β-subunits of immunoproteasomes in addition to PA200. Deletion of PA200 in mice abolishes acetylation-dependent degradation of somatic core histones during DNA double-strand breaks, and delays core histone disappearance in elongated spermatids. Purified PA200 greatly promotes ATP-independent proteasomal degradation of the acetylated core histones, but not polyubiquitinated proteins. Furthermore, acetylation on histones is required for their binding to the bromodomain-like regions in PA200 and its yeast ortholog, Blm10. Thus, PA200/Blm10 specifically targets the core histones for acetylation-mediated degradation by proteasomes, providing mechanisms by which acetylation regulates histone degradation, DNA repair, and spermatogenesis. PMID:23706739

  14. Four-fluid model of PWR degraded cores

    International Nuclear Information System (INIS)

    Dearing, J.F.

    1985-01-01

    This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

  15. Experimental investigation of coolability behaviour of irregularly shaped particulate debris bed

    International Nuclear Information System (INIS)

    Kulkarni, P.P.; Rashid, M.; Kulenovic, R.; Nayak, A.K.

    2010-01-01

    In case of a severe nuclear reactor accident, the core can melt and form a particulate debris bed in the lower plenum of the reactor pressure vessel (RPV). Due to the decay heat, the particle bed, if not cooled properly, can cause failure of the RPV. In order to avoid further propagation of the accident, complete coolability of the debris bed is necessary. For that, understanding of various phenomena taking place during the quenching is important. In the frame of the reactor safety research, fundamental experiments on the coolability of debris beds are carried out at IKE with the test facility 'DEBRIS'. In the present paper, the boiling and dry-out experimental results on a particle bed with irregularly shaped particles mixed with stainless steel balls have been reported. The pressure drops and dry-out heat fluxes of the irregular-particle bed are very similar to those for the single-sized 3 mm spheres bed, despite the fact that the irregular-particle bed is composed of particles with equivalent diameters ranging from 2 to 10 mm. Under top-flooding conditions, the pressure gradients are all smaller than the hydrostatic pressure gradient of water, indicating an important role of the counter-current interfacial drag force. For bottom-flooding with a liquid inflow velocity higher than about 2.7 mm/s, the pressure gradient generally increases consistently with the vapour velocity and the fluid-particle drag becomes important. The system pressures (1 and 3 bar) have negligible effects on qualitative behaviour of the pressure gradients. The coolability of debris beds is mainly limited by the counter-current flooding limit (CCFL) even under bottom-flooding conditions with low flow rates. The system pressure and the flow rate are found to have a distinct effect on the dry-out heat flux. Different classical models have been used to predict the pressure drop characteristics and the dry-out heat flux (DHF). Comparisons are made among the models and experimental results for

  16. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R. [Royal Inst. of Technology, Div. of Nuclear Power Safety (Sweden)

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures ({approx} 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  17. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors - EXCOOLSE project report 2004

    International Nuclear Information System (INIS)

    Park, H.S.; Nayak, A.K.; Hansson, R.C.; Sehgal, B.R.

    2005-10-01

    Beyond-the-design-basis accidents, i.e. severe accidents, involve melting of the nuclear reactor core and release of radioactivity. Intensive research has been performed for years to evaluate the consequence of the postulated severe accidents. Severe accidents posed, to the reactor researchers, a most interesting and most difficult set of phenomena to understand, and to predict the consequences, for the various scenarios that could be contemplated. The complexity of the interactions, occurring at such high temperatures (∼ 2500 deg. C), between different materials, which are changing phases and undergoing chemical reactions, is simply indescribable with the accuracy that one may desire. Thus, it is a wise approach to pursue research on SA phenomena until the remaining uncertainty in the predicted consequence, or the residual risk, can be tolerated. In the PRE-DELI-MELT project at NKS, several critical issues on the core melt loadings in the BWR and PWR reactor containments were identified. Many of Nordic nuclear power plants, particularly in boiling water reactors, adopted the Severe Accident Management Strategy (SAMS) which employed the deep subcooled water pool in lower dry-well. The success of this SAMS largely depends on the issues of steam explosions and formation of debris bed and its coolability. From the suggestions of the PRE-DELI-MELT project, a series of research plan was proposed to investigate the remaining issues specifically on the ex-vessel coolability of corium during severe accidents; (a) ex-vessel coolability of the melt or particulate debris, and (b) energetics and debris characteristics of fuel-coolant interactions endangering the integrity of the reactor containments. (au)

  18. Code development for debris bed coolability problem. Final report for the period 1997-05-01 - 1999-08-14

    International Nuclear Information System (INIS)

    Loboiko, A.I.

    2000-03-01

    The study was devoted to the problem of debris bed coolability arising from severe accident at nuclear power reactor. After reactor core melting occurs and subsequent debris bed is formed in the lower plenum of reactor pressure vessel (RPV) it is important to confine this debris bed inside RPV boundary. One of the possible accident scenarios assumes the interaction between coolant and molten core materials resulting from rapid melt quenching, freezing and fragmentation. Particulated fuel and steel may subsequently settle on available surfaces within the reactor vessel, forming debris porous beds which produce radioactive decay heating. In case of severe core degradation, such heat transfer mechanisms as radiation, conduction and natural single-phase convection may appear to be insufficient and coolant boiling may happen on the surface or inside the bed. Depending on rate of heat generation there may be sufficient debris cool down or its 'dryout' which pose a danger for RPV integrity. The study considers development of 2D numerical code capable to predict coolant saturation as a function of different parameters. Analysis of previous activities on one-dimensional and multi-dimensional models was done. On the basis of the analysis it was concluded that the correct prediction of the debris saturation on dryant power requires two-dimensional numerical simulation considering the processes like two-phase convection, capillary effects, different models of permeability, different models of heat transfer between solid debris and coolant, non-homogeneity of parameters porous medium, heat and mass transfer between debris bed and a highly porous gap along the inner RPV surface. Particular attention was given to consideration of boundary conditions for debris bed. Introduction of the analytical model for dependence of gap properties on heat flux from debris bed allowed to create an algorithm for use in numerical calculations and finally to develop a code which allowed for stable

  19. Gold Core Mesoporous Organosilica Shell Degradable Nanoparticles for Two-Photon Imaging and Gemcitabine Monophosphate Delivery

    KAUST Repository

    Rhamani, Saher; Chaix, Arnaud; Aggad, Dina; Hoang, Phuong Mai; Moosa, Basem; Garcia, Marcel; Gary-Bobo, Magali; Charnay, Clarence; Almalik, Abdulaziz; Durand, Jean-Olivier; Khashab, Niveen M.

    2017-01-01

    The synthesis of gold core degradable mesoporous organosilica shell nanoparticles is described. The nanopaticles were very efficient for two-photon luminescence imaging of cancer cells and for in vitro gemcitabine monophosphate delivery, allowing promising theranostic applications in the nanomedicine field.

  20. Gold Core Mesoporous Organosilica Shell Degradable Nanoparticles for Two-Photon Imaging and Gemcitabine Monophosphate Delivery

    KAUST Repository

    Rhamani, Saher

    2017-09-12

    The synthesis of gold core degradable mesoporous organosilica shell nanoparticles is described. The nanopaticles were very efficient for two-photon luminescence imaging of cancer cells and for in vitro gemcitabine monophosphate delivery, allowing promising theranostic applications in the nanomedicine field.

  1. Melt coolability modeling and comparison to MACE test results

    International Nuclear Information System (INIS)

    Farmer, M.T.; Sienicki, J.J.; Spencer, B.W.

    1992-01-01

    An important question in the assessment of severe accidents in light water nuclear reactors is the ability of water to quench a molten corium-concrete interaction and thereby terminate the accident progression. As part of the Melt Attack and Coolability Experiment (MACE) Program, phenomenological models of the corium quenching process are under development. The modeling approach considers both bulk cooldown and crust-limited heat transfer regimes, as well as criteria for the pool thermal hydraulic conditions which separate the two regimes. The model is then compared with results of the MACE experiments

  2. Analysis of ex-vessel melt jet breakup and coolability. Part 1: Sensitivity on model parameters and accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Moriyama, Kiyofumi; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Hwang, Byoungcheol; Jung, Woo Hyun

    2016-06-15

    Highlights: • Application of JASMINE code to melt jet breakup and coolability in APR1400 condition. • Coolability indexes for quasi steady state breakup and cooling process. • Typical case in complete breakup/solidification, film boiling quench not reached. • Significant impact of water depth and melt jet size; weak impact of model parameters. - Abstract: The breakup of a melt jet falling in a water pool and the coolability of the melt particles produced by such jet breakup are important phenomena in terms of the mitigation of severe accident consequences in light water reactors, because the molten and relocated core material is the primary heat source that governs the accident progression. We applied a modified version of the fuel–coolant interaction simulation code, JASMINE, developed at Japan Atomic Energy Agency (JAEA) to a plant scale simulation of melt jet breakup and cooling assuming an ex-vessel condition in the APR1400, a Korean advanced pressurized water reactor. Also, we examined the sensitivity on seven model parameters and five initial/boundary condition variables. The results showed that the melt cooling performance of a 6 m deep water pool in the reactor cavity is enough for removing the initial melt enthalpy for solidification, for a melt jet of 0.2 m initial diameter. The impacts of the model parameters were relatively weak and that of some of the initial/boundary condition variables, namely the water depth and melt jet diameter, were very strong. The present model indicated that a significant fraction of the melt jet is not broken up and forms a continuous melt pool on the containment floor in cases with a large melt jet diameter, 0.5 m, or a shallow water pool depth, ≤3 m.

  3. In-vessel coolability and steam explosion in Nordic BWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ma, W.; Li, L.; Hansson, R.; Villanueva, W.; Kudinov, P.; Manickam, L.; Tran, C.-T. (Royal Institute of Technology (KTH) (Sweden))

    2011-05-15

    The objective of this research is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in the Nordic BWR plants which employ cavity flooding as severe accident management (SAM) strategy. To quantify the coolability of debris bed packed with irregular particles, the friction laws of fluid flow in particulate beds packed with non-spherical particles were investigated on the POMECO-FL test facility, and the experimental data suggest that the Ergun equation is applicable if the effective particle diameter of the particles is represented by the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles. One-way coupling analysis between PECM model for melt pool heat transfer and ANSYS thermo-structural mechanics was performed to analyze the vessel creep, and the results revealed two different modes of vessel failure: a 'ballooning' of the vessel bottom and a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Single-droplet steam explosion experiments were carried out by using oxidic mixture of WO{sub 3}-CaO, and the results show an apparent difference in steam explosion energetics between the eutectic and non-eutectic melts at low melt superheat (100 deg. C). (Author)

  4. Crust formation and its effect on the molten pool coolability

    Energy Technology Data Exchange (ETDEWEB)

    Park, R.J.; Lee, S.J.; Sim, S.K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-09-01

    Experimental and analytical studies of the crust formation and its effect on the molten pool coolability have been performed to examine the crust formation process as a function of boundary temperatures as well as to investigate heat transfer characteristics between molten pool and overlying water in order to evaluate coolability of the molten pool. The experimental test results have shown that the surface temperature of the bottom plate is a dominant parameter in the crust formation process of the molten pool. It is also found that the crust thickness of the case with direct coolant injection into the molten pool is greater than that of the case with a heat exchanger. Increasing mass flow rate of direct coolant injection to the molten pool does not affect the temperature of molten pool after the crust has been formed in the molten pool because the crust behaves as a thermal barrier. The Nusselt number between the molten pool and the coolant of the case with no crust formation is greater than that of the case with crust formation. The results of FLOW-3D analyses have shown that the temperature distribution contributes to the crust formation process due to Rayleigh-Benard natural convection flow.

  5. In-vessel coolability and steam explosion in Nordic BWRs

    International Nuclear Information System (INIS)

    Ma, W.; Li, L.; Hansson, R.; Villanueva, W.; Kudinov, P.; Manickam, L.; Tran, C.-T.

    2011-05-01

    The objective of this research is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in the Nordic BWR plants which employ cavity flooding as severe accident management (SAM) strategy. To quantify the coolability of debris bed packed with irregular particles, the friction laws of fluid flow in particulate beds packed with non-spherical particles were investigated on the POMECO-FL test facility, and the experimental data suggest that the Ergun equation is applicable if the effective particle diameter of the particles is represented by the equivalent diameter of the particles, which is the product of Sauter mean diameter and shape factor of the particles. One-way coupling analysis between PECM model for melt pool heat transfer and ANSYS thermo-structural mechanics was performed to analyze the vessel creep, and the results revealed two different modes of vessel failure: a 'ballooning' of the vessel bottom and a 'localized creep' concentrated within the vicinity of the top surface of the melt pool. Single-droplet steam explosion experiments were carried out by using oxidic mixture of WO 3 -CaO, and the results show an apparent difference in steam explosion energetics between the eutectic and non-eutectic melts at low melt superheat (100 deg. C). (Author)

  6. Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

    1995-09-01

    The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

  7. Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

    International Nuclear Information System (INIS)

    Chung, H.M.; Thomas, G.R.

    1983-02-01

    Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850 0 C or extrapolated from the low-temperature data obtained at 0 C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, rather than in an unlimited flux of pure steam. New results of Zircaloy oxidation measurements in various mixtures of hydrogen and steam are reported for >1500 0 C. The results show significantly smaller oxidation and, hence, hydrogen-generation rates in the mixture, compared with those obtained in pure steam. It is also shown that a significant fraction of hydrogen, generated as a result of Zircaloy oxidation, is dissolved in the cladding material itself, which prevents that portion of hydrogen from reaching the containment building space. Implications of these findings are discussed in relation to a more realistic method of quantifying the hydrogen source term for a degraded-core accident analysis

  8. In-vessel coolability and steam explosion in Nordic BWRs

    International Nuclear Information System (INIS)

    Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T.

    2010-05-01

    The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixtures properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (∼10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower superheat. (author)

  9. A critical role for protein degradation in the nucleus accumbens core in cocaine reward memory.

    Science.gov (United States)

    Ren, Zhen-Yu; Liu, Meng-Meng; Xue, Yan-Xue; Ding, Zeng-Bo; Xue, Li-Fen; Zhai, Suo-Di; Lu, Lin

    2013-04-01

    The intense associative memories that develop between cocaine-paired contexts and rewarding stimuli contribute to cocaine seeking and relapse. Previous studies have shown impairment in cocaine reward memories by manipulating a labile state induced by memory retrieval, but the mechanisms that underlie the destabilization of cocaine reward memory are unknown. In this study, using a Pavlovian cocaine-induced conditioned place preference (CPP) procedure in rats, we tested the contribution of ubiquitin-proteasome system-dependent protein degradation in destabilization of cocaine reward memory. First, we found that polyubiquitinated protein expression levels and polyubiquitinated N-ethylmaleimide-sensitive fusion (NSF) markedly increased 15 min after retrieval while NSF protein levels decreased 1 h after retrieval in the synaptosomal membrane fraction in the nucleus accumbens (NAc) core. We then found that infusion of the proteasome inhibitor lactacystin into the NAc core prevented the impairment of memory reconsolidation induced by the protein synthesis inhibitor anisomycin and reversed the effects of anisomycin on NSF and glutamate receptor 2 (GluR2) protein levels in the synaptosomal membrane fraction in the NAc core. We also found that lactacystin infusion into the NAc core but not into the shell immediately after extinction training sessions inhibited CPP extinction and reversed the extinction training-induced decrease in NSF and GluR2 in the synaptosomal membrane fraction in the NAc core. Finally, infusions of lactacystin by itself into the NAc core immediately after each training session or before the CPP retrieval test had no effect on the consolidation and retrieval of cocaine reward memory. These findings suggest that ubiquitin-proteasome system-dependent protein degradation is critical for retrieval-induced memory destabilization.

  10. Experimental results on the coolability of a debris bed with multidimensional cooling effects

    International Nuclear Information System (INIS)

    Rashid, M.; Kulenovic, R.; Laurien, E.; Nayak, A.K.

    2011-01-01

    Research highlights: ► Performing of dryout experiments with a polydispersed bed for top- and bottom-flooding. ► Study of influence of different down comer configurations on the coolability of debris bed. ► Measurement of temperature profiles, pressure drops and determination of dryout heat flux. ► Observation of noticeable increase in coolability of debris bed with the use of down comer is observed. - Abstract: Within the reactor safety research, the removal of decay heat from a debris bed (formed from corium and residual water) is of great importance. In order to investigate experimentally the long term coolability of debris beds, the scaled test facility “DEBRIS” (Fig. 1) has been built at IKE. A large number of experiments had been carried out to investigate the coolability limits for different bed configurations (). Analyses based on one-dimensional configurations underestimate the coolability in realistic multidimensional configurations, where lateral water access and water inflow via bottom regions are favoured. Following the experiments with top- and bottom-flooding flow conditions this paper presents experimental results of boiling and dryout tests at different system pressures based on top- and bottom-flooding via a down comer configuration. A down comer with an internal diameter of 10 mm has been installed at the centre of the debris bed. The debris bed is built up in a cylindrical crucible with an inner diameter of 125 mm. The bed of height 640 mm is composed of polydispersed particles with particle diameters 2, 3 and 6 mm. Since the long term coolability of such particle bed is limited by the availability of coolant inside the bed and not by heat transfer limitations from the particles to the coolant, the bottom inflow of water improves the coolability of the debris bed and an increase of the dryout heat flux can be observed. With increasing system pressure, the coolability limits are enhanced (increased dryout heat flux).

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

    CERN Document Server

    Kelly, G N

    1983-01-01

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

  12. Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

    International Nuclear Information System (INIS)

    Kim, S.S.; Levine, S.H.

    1985-01-01

    An analytical/experimental method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs nucleonic codes to analyze the neutron multiplication of a Cf-252 neutron source. An optimization program, GPM, is utilized to unfold the k/sub infinity/ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn State Breazeale TRIGA Reactor (PSBR). A significant result of this study is that it provides a method to monitor the criticality of a damaged core during the recovery period

  13. In-Vessel Coolability. Workshop Proceedings, in collaboration with EC-SARNET

    International Nuclear Information System (INIS)

    2011-01-01

    Severe Accident Management Guidelines increase focus on containment integrity after some progression in the course of a severe accident. This change in priorities is made according to criteria that vary depending on reactor type and specific procedures. Once a water source has been recovered, different accident management strategies can be used: send water into the core and/or cool the reactor pressure vessel (RPV) externally. It should be noticed that, depending on the amount of water available, these strategies might conflict with other uses of water such as for instance activating spray systems in the containment or may have deleterious effects as for instance an increase in the production of hydrogen. Generally, for in-vessel reflooding, the models used for evaluation of accident management measures suffer from a lack of validation. Given this background, the objectives of the workshop were: -) to exchange information on different Severe Accident Management strategies used or contemplated for the in-vessel coolability issue; -) to review recent, ongoing and planned experimental programmes on reflooding; -) to review models used for reflooding in severe accident calculation tools, either simplified or sophisticated; -) to exchange information on the treatment of reflooding in different safety studies such as Probabilistic Safety Assessment; and -) to provide recommendations for future work, as necessary

  14. Experimental studies on the coolability of packed beds. Flooding of hot dry packed beds

    International Nuclear Information System (INIS)

    Leininger, S.; Kulenovic, R.; Laurien, E.

    2013-01-01

    In case of a severe accident in a nuclear power plant meltdown of the reactor core can occur and form a packed bed in the lower plenum of the reactor pressure vessel (RPV) after solidification due to contact with water. The removal of after-heat and the long-term coolability is of essential interest. The efficient injection of cooling water into the packed bed has to be assured without endangering the structural integrity of the reactor pressure vessel. The experiments performed aimed to study the dry-out and the quenching (flooding) of hot dry packed beds. Two different inflow variants, bottom- and top-flooding including the variation of the starting temperature of the packed bed and the injection rate were studied. In case of bottom flooding the quenching time increases with increasing packed bed temperature and decreasing injection rate. In case of top flooding the flow pattern is more complex, in a first phase the water flows preferentially toward the RPV wall, the flow paths conduct the water downwards. The flow resistance of the packed bed increases with increasing bed temperatures. The quenching temperatures increase significantly above average.

  15. Degradation of austenitic stainless steel (SS) light water ractor (LWR) core internals due to neutron irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Appajosula S., E-mail: Appajosula.Rao@nrc.gov

    2014-04-01

    Austenitic stainless steels (SSs) are extensively being used in the fabrication of light water reactor (LWR) core internal components. It is because these steels have relatively high ductility, fracture toughness and moderate strength. However, the LWR internal components exposure to neutron irradiation over an extended period of plant operation degrades the materials mechanical properties such as the fracture toughness. This paper summarizes some of the results of the existing open literature data on irradiation assisted stress corrosion cracking (IASCC) of 316 CW steels that have been published by the United States Nuclear Regulatory Commission (USNRC), industry, academia, and other research agencies.

  16. Simulation of the PHEBUS FPT-1 experiment using MELCOR and exploration of the primary core degradation mechanism

    International Nuclear Information System (INIS)

    Wang, Jun; Corradini, Michael L.; Fu, Wen; Haskin, Troy; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Core degradation evaluation is an important process in risk analysis. • PHEBUS experiment was simulated using MELCOR. • The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. • These results are used to analyze the mode and behavior of core degradation. - Abstract: Core degradation evaluation of probability, progression and consequences of a core degradation accident is critical for evaluation of risk as well as its mitigation. However, research and modeling of severe accidents to date are limited, and their accuracy in predicting severe accident consequences is still insufficient. It is therefore important to explore the mechanisms of core degradation and to develop mitigation measures for severe accidents. PHEBUS FPT1 is a typical and classic core degradation experiment. MELCOR is a world famous severe accident analysis code developed by Sandia National Lab that has seen wide application, a broad user base, and a number of supporting experiments. The PHEBUS experiment was simulated using MELCOR in this paper. Experimental data on, thermal power and steam mass flow rates are used to determine average pressure, energy distribution, molten mass, temperature of the fuel, and hydrogen generation. Data from the PHEBUS experiment and Cho’s calculations are used to compare the average pressure, several fuel temperatures and the hydrogen generation rate. The results confirm the validity of MELCOR’s simulation of the PHEBUS experiment. The temperature distribution of the core is provided. These results are used to determine the mode and behavior of core degradation with the intent of building a foundation for further research

  17. Comparison of CORA and MELCOR core degradation simulation and the MELCOR oxidation model

    International Nuclear Information System (INIS)

    Wang, Jun; Corradini, Michael L.; Fu, Wen; Haskin, Troy; Tian, Wenxi; Zhang, Yapei; Su, Guanghui; Qiu, Suizheng

    2014-01-01

    Highlights: • Oxidation model of MELCOR is analyzed and the improving suggestion is provided. • MELCOR core degradation calculating results are compared with CORA experiment. • Flow rate of argon and steam, the generating rate of hydrogen is calculated and compared. • Temperature spatial variation and temperature history is calculated and presented. - Abstract: MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin–Madison under the help of Sandia. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed

  18. Comparison of CORA and MELCOR core degradation simulation and the MELCOR oxidation model

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jun [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China); Corradini, Michael L., E-mail: corradini@engr.wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); Fu, Wen [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Haskin, Troy [College of Engineering, The University of Wisconsin-Madison, Madison, WI 53706 (United States); Tian, Wenxi; Zhang, Yapei; Su, Guanghui; Qiu, Suizheng [State Key Laboratory of Multiphase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 (China)

    2014-09-15

    Highlights: • Oxidation model of MELCOR is analyzed and the improving suggestion is provided. • MELCOR core degradation calculating results are compared with CORA experiment. • Flow rate of argon and steam, the generating rate of hydrogen is calculated and compared. • Temperature spatial variation and temperature history is calculated and presented. - Abstract: MELCOR is widely used and sufficiently trusted for severe accident analysis. However, the occurrence of Fukushima has increased the focus on severe accident codes and their use. A MELCOR core degradation calculation was conducted at the University of Wisconsin–Madison under the help of Sandia. The calculation results were checked by comparing with a past CORA experiment. MELCOR calculation results included the flow rate of argon and steam, the generation rate of hydrogen. Through this work, the performance of MELCOR COR package was reviewed in detail. This paper compares the hydrogen generation rates predicted by MELCOR to the CORA test data. While agreement is reasonable it could be improved. Additionally, the MELCOR zirconium oxidation model was analyzed.

  19. In-vessel core degradation in LWR severe accidents: a state of the art report to CSNI january 1991

    International Nuclear Information System (INIS)

    1991-11-01

    This state of the art report on in-vessel core degradation has been produced at the request of CSNI Principal Working Group 2. The objective of the report is to present to CSNI member countries the status of research and related information on in-vessel degraded core behaviour in both Pressurised Water Reactors (PWR) and Boiling Water Reactors (BWR). Information on experiments, codes and comparisons of calculations with experiments up to january 1991 is summarised and reviewed. Integrated codes, which are wider in scope than just in-vessel degradation are covered as well as specialist, degraded core codes. Implications for PWR and BWR plant calculations are considered. Conclusions and recommendations for research, plant calculations and further CSNI activity in this area are the subject of the final chapter. A major conclusion of the report is that early phase core degradation is relatively well understood. However, codes need further development to bring them up to date with the experimental database, particularly to include low temperature liquefaction processes. These processes significantly affect early phase core degradation and their neglect could affect assessments of accident management actions (including recriticality in BWR severe accidents)

  20. An experimental study on coolability of a particulate bed with radial stratification or triangular shape

    International Nuclear Information System (INIS)

    Thakre, Sachin; Li, Liangxing; Ma, Weimin

    2014-01-01

    Highlights: • Dryout heat flux of a particulate bed with radial stratification is obtained. • It was found to be dominated by hydrodynamics in the bigger size of particle layer. • Coolability of a particulate bed with triangular shape is investigated. • The coolability is improved in the triangular bed due to lateral ingression of coolant. • Coolability of both beds is enhanced by a downcomer. - Abstract: This paper deals with the results of an experimental study on the coolability of particulate beds with radial stratification and triangular shape, respectively. The study is intended to get an idea on how the coolability is affected by the different features of a debris bed formed in a severe accident of light water reactors. The experiments were performed on the POMECO-HT facility which was constructed to investigate two-phase flow and heat transfer in particulate beds under either top-flooding or bottom-fed condition. A downcomer is designed to enable investigation of the effectiveness of natural circulation driven coolability. Two homogenous beds were also employed in the present study to compare their dryout power densities with those of the radially stratified bed and the triangular bed. The results show that the dryout heat fluxes of the homogeneous beds at top-flooding condition can be predicted by the Reed model. For the radially stratified bed, the dryout heat flux is dominated by two-phase flow in the columns packed with larger particles, and the dryout occurred initially near the boundary between the middle column and a side column. Given the same volume of particles under top-flooding condition, the dryout power density of the triangular bed is about 69% higher than that of the homogenous bed. The coolability of all the beds is enhanced by bottom-fed coolant driven by either forced injection or downcomer-induced natural circulation

  1. An experimental study on coolability of a particulate bed with radial stratification or triangular shape

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, Sachin; Li, Liangxing; Ma, Weimin, E-mail: ma@safety.sci.kth.se

    2014-09-15

    Highlights: • Dryout heat flux of a particulate bed with radial stratification is obtained. • It was found to be dominated by hydrodynamics in the bigger size of particle layer. • Coolability of a particulate bed with triangular shape is investigated. • The coolability is improved in the triangular bed due to lateral ingression of coolant. • Coolability of both beds is enhanced by a downcomer. - Abstract: This paper deals with the results of an experimental study on the coolability of particulate beds with radial stratification and triangular shape, respectively. The study is intended to get an idea on how the coolability is affected by the different features of a debris bed formed in a severe accident of light water reactors. The experiments were performed on the POMECO-HT facility which was constructed to investigate two-phase flow and heat transfer in particulate beds under either top-flooding or bottom-fed condition. A downcomer is designed to enable investigation of the effectiveness of natural circulation driven coolability. Two homogenous beds were also employed in the present study to compare their dryout power densities with those of the radially stratified bed and the triangular bed. The results show that the dryout heat fluxes of the homogeneous beds at top-flooding condition can be predicted by the Reed model. For the radially stratified bed, the dryout heat flux is dominated by two-phase flow in the columns packed with larger particles, and the dryout occurred initially near the boundary between the middle column and a side column. Given the same volume of particles under top-flooding condition, the dryout power density of the triangular bed is about 69% higher than that of the homogenous bed. The coolability of all the beds is enhanced by bottom-fed coolant driven by either forced injection or downcomer-induced natural circulation.

  2. Analyses on ex-vessel debris formation and coolability in SARNET frame

    International Nuclear Information System (INIS)

    Pohlner, G.; Buck, M.; Meignen, R.; Kudinov, P.; Ma, W.; Polidoro, F.; Takasuo, E.

    2014-01-01

    Highlights: • Melt outflow varies from dripping melt outflow to molten corium jets of variable size. • Experiments show clear trend of producing particles in size range 2-4 mm. • Code calculations show complete solidification of particles, yielding formation of fragmented debris beds. • Limits of debris bed cooling and coolability margins are analysed. - Abstract: The major aim of work in the SARNET2 European project on ex-vessel debris formation and coolability was to get an overall perspective on coolability of melt released from a failed reactor pressure vessel and falling into a water-filled cavity. Especially, accident management concepts for BWRs, dealing with deep water pools below the reactor vessel, are addressed, but also shallower pools in existing PWRs, with questions about partial cooling and time delay of molten corium concrete interaction. The subject can be divided into three main topics: (i) Debris bed formation by breakup of melt, (ii) Coolability of debris and (iii) Coupled treatment of the processes. Accompanied by joint collaborations of the partners, the performed work comprises theoretical, experimental and modelling activities. Theoretical work was done by KTH on the melt outflow conditions from a RPV and on the quantification of the probability of yielding a non-coolable ex-vessel bed by use of probabilistic assessment. IKE introduced a theoretical concept to improve debris bed coolability. A large amount of experimental work was done by partners (KTH, VTT, IKE) on the coolability of debris beds using different bed geometries, particles, heating methods and water feeds, yielding a valuable base for code validation. Modelling work was mainly done by IKE, IRSN, RSE and VTT concerning jet breakup and/or debris bed formation and cooling in 2D and 3D geometries. A benchmark for the DEFOR-A experiment of KTH was performed. Important progress was reached for several tasks and aspects and important insights are given, enabling to focus the

  3. Generic BWR-4 degraded core in-vessel study. Status report

    International Nuclear Information System (INIS)

    1984-11-01

    Original intent of this project was to produce a phenomenological study of the in-vessel degradation which occurs during the TQUX and TQUV sequences for a generic BWR-4 from the initiation of the FSAR Chapter 15 operational transient through core debris bed formation to the failure of the primary pressure boundary. Bounding calculations were to be performed for the two high pressure and low pressure non-LOCA scenarios to assess the uncertainties in the current state of knowledge regarding the source terms for containment integrity studies. Source terms as such were defined in terms of hydrogen generation, unreacted metal, and coolant inventroy, and in terms of the form, sequencing and mode of dispersal through the primary vessel boundary. Fission product release was not to be considered as part of this study. Premature termination of the project, however, led to the dicontinuation of work on an as is basis. Work on the in-core phase from the point of scram to core debris bed formation was largely completed. A preliminary scoping calculation on the debris bed phase had been initiated. This report documents the status of the study at termination

  4. Experimental investigations on the coolability of prototypical particle beds with respect to reactor safety; Experimentelle Untersuchungen der Kuehlbarkeit prototypischer Schuettungskonfigurationen unter dem Aspekt der Reaktorsicherheit

    Energy Technology Data Exchange (ETDEWEB)

    Leininger, Simon

    2017-02-22

    In case of a severe accident in a light water reactor, continuous unavailability of cooling water to the reactor core may result in overheating of the fuel elements and finally the loss of core integrity. Under such conditions, a structure of heat-releasing particles of different size and shape may be formed by fragmentation of molten core material in several stages of the accident. The long-term coolability of such beds is of prime im-portance to avoid any damage to the reactor pressure vessel or even a release of fission products to the environment. In the frame of this work, specific experiments were con-ducted under prototypical conditions employing the existing DEBRIS test facility in order to gain further knowledge about the thermohydraulic behavior of such beds. In steady state boiling experiments, the pressure gradients in particle beds were meas-ured both for one- and multi-dimensional cooling water flow conditions and compared with one another in order to assess the flow behavior inside the bed. For these different flow conditions as well as for stratified bed configurations, the maximum removable heat flux densities were determined in the dryout experiments. E. g., it was found that an axial stratification of the permeability can significantly reduce the bed's coolability. For the first time, the quenching behavior of dry, superheated beds was investigated at elevated system pressure up to 0.5 MPa. In these experiments, the effect of system pressure on the coolability was quantified by means of the quenching time (time period to cool down the bed to saturation temperature). The investigated particle beds mainly consisted of non-spherical particles with well-defined geometry (cylinders and screws). It was shown that the effect of the particles geometry on the flow in a particle bed can be best estimated by using an equivalent particle diameter calculated for monodisperse particle beds from the product of the Sauter diameter and a shape factor and for

  5. Degraded core accidents for the Sizewell PWR A sensitivity analysis of the radiological consequences

    CERN Document Server

    Kelly, G N; Clarke, R H; Ferguson, L; Haywood, S M; Hemming, C R; Jones, J A

    1982-01-01

    The radiological impact of degraded core accidents postulated for the Sizewell PWR was assessed in an earlier study. In this report the sensitivity of the predicted consequences to variation in the values of a number of important parameters is investigated for one of the postulated accidental releases. The parameters subjected to sensitivity analyses are the dose-mortality relationship for bone marrow irradiation, the energy content of the release, the warning time before the release to the environment, and the dry deposition velocity for airborne material. These parameters were identified as among the more important in determining the uncertainty in the results obtained in the initial study. With a few exceptions the predicted consequences were found to be not very sensitive to the parameter values investigated, the range of variation in the consequences for the limiting values of each parameter rarely exceeded a factor of a few and in many cases was considerably less. The conclusions reached are, however, p...

  6. The effect of self-leveling on debris bed coolability under severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Basso, S.; Konovalenko, A. [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, D5, Stockholm 106 91 (Sweden); Yakush, S.E. [Institute for Problems in Mechanics of the Russian Academy of Sciences, Ave. Vernadskogo 101 Bldg 1, Moscow 119526 (Russian Federation); Kudinov, P. [Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, D5, Stockholm 106 91 (Sweden)

    2016-08-15

    Highlights: • A model for coolability of a self-leveling, variable-shape debris bed is proposed. • Sensitivity analysis is performed to screen out the less influential input parameters. • A small fraction of scenarios has initially a non-coolable debris bed configuration. • The fraction of non-coolable scenarios decreases with time due to self-leveling. - Abstract: Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWRs, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the

  7. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  8. Degraded Core Quench: Summary of Progress 1996-1999 - Executive Summary

    International Nuclear Information System (INIS)

    Haste, T.J.; Trambauer, K.

    2000-01-01

    A status report on experiments and modelling relating to quench of degraded cores was issued by CSNI in August 1996, following the publication of the In-Vessel Core Degradation Code Validation Matrix. In response to a request by PWG2 through the TG-DCC, a review of progress since then to June 1999 has been performed. The scope is broadly the same as before, restricted to mainly rod-like geometries and not considering pure debris bed configurations. The scope has been increased slightly to include a VVER bundle quench experiment, CODEX-3, which falls within the parameter range of the Western bundle experiments performed to date. The same format has been adopted as before, with the experimental results for bundle and separate-effects tests being summarised in separate tables, updated from the earlier report. This review shows further evolutionary progress made in understanding the phenomena of fuel rod quench under severe accident conditions. The successful performance of commissioning and four main tests in the new bundle QUENCH facility at FZ Karlsruhe has provided valuable new data, supplemented by the VVER test CODEX-3 at AEKI Budapest. Temperature excursions and excess hydrogen production were only observed for quench from high temperature (2300 K) with a non pre-oxidised bundle (2 relevant tests); for quench from lower temperatures (1750-1870 K) and with pre-oxidation (50- 500 μm oxide) smooth cooling with no significant excess hydrogen production was observed (3 relevant tests). When cooling a non pre-oxidised bundle from 1870 K rapidly by steam, no significant excursion was observed (1 test). These new lower temperature bundle tests have usefully extended the parameter range down from that previously covered (quench temperature 2150 K and above, no pre-oxidation, temperature excursions/excess hydrogen production always observed), and have shown that there are conditions for quench from high temperature where excess temperatures and hydrogen production do not

  9. Oxidative degradation of the antibiotic oxytetracycline by Cu@Fe3O4 core-shell nanoparticles.

    Science.gov (United States)

    Pham, Van Luan; Kim, Do-Gun; Ko, Seok-Oh

    2018-08-01

    A core-shell nanostructure composed of zero-valent Cu (core) and Fe 3 O 4 (shell) (Cu@Fe 3 O 4 ) was prepared by a simple reduction method and was evaluated for the degradation of oxytetracycline (OTC), an antibiotic. The Cu core and the Fe 3 O 4 shell were verified by X-ray diffractometry (XRD) and transmission electron microscopy. The optimal molar ratio of [Cu]/[Fe] (1/1) in Cu@Fe 3 O 4 created an outstanding synergic effect, leading to >99% OTC degradation as well as H 2 O 2 decomposition within 10min at the reaction conditions of 1g/L Cu@Fe 3 O 4 , 20mg/L OTC, 20mM H 2 O 2 , and pH3.0 (and even at pH9.0). The OTC degradation rate by Cu@Fe 3 O 4 was higher than obtained using single nanoparticle of Cu or Fe 3 O 4 . The results of the study using radical scavengers showed that OH is the major reactive oxygen species contributing to the OTC degradation. Finally, good stability, reusability, and magnetic separation were obtained with approximately 97% OTC degradation and no notable change in XRD patterns after the Cu@Fe 3 O 4 catalyst was reused five times. These results demonstrate that Cu@Fe 3 O 4 is a novel prospective candidate for the pharmaceutical and personal care products degradation in the aqueous phase. Copyright © 2018 Elsevier B.V. All rights reserved.

  10. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Dinh, T.N. [Royal Institute of Technology (Sweden)

    2007-04-15

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  11. Ex-vessel coolability and energetics of steam explosions in nordic light water reactors

    International Nuclear Information System (INIS)

    Park, H.S.; Dinh, T.N.

    2007-04-01

    The report summarizes activities conducted at the Division of Nuclear Power Safety, Royal Institute of Technology-Sweden (KTH-NPS) within the ExCoolSe project during the year 2005, which is a transition year for the KTH-NPS program. The ExCoolSe project supported by NKS contributes to the severe accident research at KTH-NPS concurrently supported by APRI, HSK and EU SARNET. The main objective in ExCoolSe project is to scrutinize research on risk-significant safety issues related to severe accident management (SAM) strategy adopted for Nordic BWR plants, namely the Ex-vessel Coolability and Energetic Steam explosion. The work aims to pave way toward building a tangible research framework to tackle these long-standing safety issues. Chapter 1 describes the project objectives and work description. Chapter 2 provides a critical assessment of research results obtained from several past programs at KTH. This includes review of key data, insights and implications from POMECO (Porous Media Coolability) program, COMECO (Corium Melt Coolability) program, SIMECO (Study of In-Vessel Melt Coolability) program, and MISTEE (Micro-Interactions in Steam Explosion Experiments) program. Chapter 3 discusses the rationale of the new research program focusing on the SAM issue resolution. The program emphasizes identification and qualification of physics-based limiting mechanisms for both in-vessel phenomena (melt progression and debris coolability in the lower head, vessel failure), and ex-vessel phenomena. Chapter 4 introduces research results from the newly established DEFOR (Debris Formation) program and the ongoing MISTEE program. The focus of DEFOR is fulfill an apparent gap in the contemporary knowledge of severe accidents, namely mechanisms which govern the debris bed formation and bed characteristics. The later control the debris bed coolability. In the MISTEE program, methods for image synchronization and data processing were developed and tested, which enable processing of

  12. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  13. Pt@Ag and Pd@Ag core/shell nanoparticles for catalytic degradation of Congo red in aqueous solution

    Science.gov (United States)

    Salem, Mohamed A.; Bakr, Eman A.; El-Attar, Heba G.

    2018-01-01

    Platinum/silver (Pt@Ag) and palladium/silver (Pd@Ag) core/shell NPs have been synthesized in two steps reaction using the citrate method. The progress of nanoparticle formation was followed by the UV/Vis spectroscopy. Transmission electron microscopy revealed spherical shaped core/shell nanoparticles with average particle diameter 32.17 nm for Pt@Ag and 8.8 nm for Pd@Ag. The core/shell NPs were further characterized by FT-IR and XRD. Reductive degradation of the Congo red dye was chosen to demonstrate the excellent catalytic activity of these core/shell nanostructures. The nanocatalysts act as electron mediators for the transfer of electrons from the reducing agent (NaBH4) to the dye molecules. Effect of reaction parameters such as nanocatalyst dose, dye and NaBH4 concentrations on the dye degradation was investigated. A comparison between the catalytic activities of both nanocatalysts was made to realize which of them the best in catalytic performance. Pd@Ag was the higher in catalytic activity over Pt@Ag. Such greater activity is originated from the smaller particle size and larger surface area. Pd@Ag nanocatalyst was catalytically stable through four subsequent reaction runs under the utilized reaction conditions. These findings can thus be considered as possible economical alternative for environmental safety against water pollution by dyes.

  14. TMI-2 core examination

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  15. Application of CAMP code to analysis of debris coolability experiments in ALPHA program

    International Nuclear Information System (INIS)

    Maruyama, Yu; Moriyama, Kiyofumi; Park, Hyun-Sun; Yang, Yanhua; Sugimoto, Jun

    1999-01-01

    An analytical code for thermo-fluid dynamics of a molten debris, CAMP, was applied to the analysis of the ex-vessel and in-vessel debris coolability experiments performed in ALPHA program. The analysis on the ex-vessel debris coolability experiments, where water was added onto a layer of thermite melt, indicated that the upper surface of the melt was remained molten during a period when melt eruptions followed by a mild steam explosion were observed. This might imply that a coarse mixing between the melt and the overlying water could have been formed if a sufficient force was generated at the interface between the two fluids. In the analysis of the in-vessel debris coolability experiments, where an aluminum oxide (Al 2 O 3 ) melt was poured into a water-filled lower head experimental vessel, a temperature increase at the outer surface of the vessel was qualitatively reproduced when a gap was assumed to be at the interface between the solidified Al 2 O 3 and the vessel wall. (author)

  16. Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

    1996-01-01

    The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

  17. An AgI@g-C3N4 hybrid core@shell structure: Stable and enhanced photocatalytic degradation

    Science.gov (United States)

    Liu, Li; Qi, Yuehong; Yang, Jinyi; Cui, Wenquan; Li, Xingang; Zhang, Zisheng

    2015-12-01

    A novel visible-light-active material AgI@g-C3N4 was prepared by ultrasonication/chemisorption method. The core@shell structure AgI@g-C3N4 catalyst showed high efficiency for the degradation of MB under visible light irradiation (λ > 420 nm). Nearly 96.5% of MB was degraded after 120 min of irradiation in the presence of the AgI@g-C3N4 photocatalyst. Superior stability was also observed in the cyclic runs indicating that the as prepared hybrid composite is highly desirable for the remediation of organic contaminated wastewaters. The improved photocatalytic performance is due to synergistic effects at the interface of AgI and g-C3N4 which can effectively accelerate the charge separation and reinforce the photostability of hybrid composite. The possible mechanism for the photocatalytic activity of AgI@g-C3N4 was tentatively proposed.

  18. BNL program in support of LWR degraded-core accident analysis

    International Nuclear Information System (INIS)

    Ginsberg, T.; Greene, G.A.

    1982-01-01

    Two major sources of loading on dry watr reactor containments are steam generatin from core debris water thermal interactions and molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON (Muir, 1981) and MARCH (Wooton, 1980) computer codes. Progress in the two programs is described in this paper. 8 figures

  19. Application of Ni-Oxide@TiO₂ Core-Shell Structures to Photocatalytic Mixed Dye Degradation, CO Oxidation, and Supercapacitors.

    Science.gov (United States)

    Lee, Seungwon; Lee, Jisuk; Nam, Kyusuk; Shin, Weon Gyu; Sohn, Youngku

    2016-12-20

    Performing diverse application tests on synthesized metal oxides is critical for identifying suitable application areas based on the material performances. In the present study, Ni-oxide@TiO₂ core-shell materials were synthesized and applied to photocatalytic mixed dye (methyl orange + rhodamine + methylene blue) degradation under ultraviolet (UV) and visible lights, CO oxidation, and supercapacitors. Their physicochemical properties were examined by field-emission scanning electron microscopy, X-ray diffraction analysis, Fourier-transform infrared spectroscopy, and UV-visible absorption spectroscopy. It was shown that their performances were highly dependent on the morphology, thermal treatment procedure, and TiO₂ overlayer coating.

  20. Status Report on Ex-Vessel Coolability and Water Management

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Robb, K. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-09-15

    Specific to BWR plants, current accident management guidance calls for flooding the drywell to a level of approximately 1.2 m (4 feet) above the drywell floor once vessel breach has been determined. While this action can help to submerge ex-vessel core debris, it can also result in flooding the wetwell and thereby rendering the wetwell vent path unavailable. An alternate strategy is being developed in the industry guidance for responding to the severe accident capable vent Order, EA-13-109. The alternate strategy being proposed would throttle the flooding rate to achieve a stable wetwell water level while preserving the wetwell vent path. The overall objective of this work is to upgrade existing analytical tools (i.e. MELTSPREAD and CORQUENCH - which have been used as part of the DOE-sponsored Fukushima accident analyses) in order to provide flexible, analytically capable, and validated models to support the development of water throttling strategies for BWRs that are aimed at keeping ex-vessel core debris covered with water while preserving the wetwell vent path.

  1. 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.

  2. Refined model for the coolability of core debris with flow entry from the bottom

    International Nuclear Information System (INIS)

    Schulenberg, T.; Mueller, U.

    1986-01-01

    Within the context of a hypothetical severe accident in light water reactors also heat generating debris beds of a coarse particle size are discussed. A refined model for two-phase flow in particle beds is presented. Compared to previous models this model takes into account the effect of interfacial drag forces between liquid and vapor. These effects are important in coarse debris beds. The model is based on the momentum equations for separated flow, which are closed by empirical relations for the wall shear stress and the interfacial drag. When the refined model is applied to LWR severe accident scenarios an increased dryout heat flux is predicted for debris beds with flow entry from the bottom driven by a moderate downcomer head

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

    Henrie, J.O.

    1988-01-01

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

  6. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    International Nuclear Information System (INIS)

    Hozer, Z.; Nagy, I.; Windberg, P.; Balasko, M.; Matus, L.; Prokopiev, O.; Pinter, A.; Horvath, M.; Gyenes, Gy.; Czitrovszky, A.; Nagy, A.; Jani, P.

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25 th May 2001 in the framework of the COLOSS project of the EU 5 th FWP. The high temperature degradation of a VVER-1000 type bundle with B 4 C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures ∼2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  7. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z; Nagy, I; Windberg, P; Balasko, M; Matus, L; Prokopiev, O; Pinter, A; Horvath, M; Gyenes, Gy [KFKI Atomic Energy Research Institute, Budapest (Hungary); Czitrovszky, A; Nagy, A; Jani, P [Research Institute for Solid State Physics and Optics, Budapest (Hungary)

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25{sup th} May 2001 in the framework of the COLOSS project of the EU 5{sup th} FWP. The high temperature degradation of a VVER-1000 type bundle with B{sub 4}C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures {approx}2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  8. Synthesis and characterization of CdS/CuAl2O4 core-shell: application to photocatalytic eosin degradation

    Science.gov (United States)

    Bellal, B.; Trari, M.; Afalfiz, A.

    2015-08-01

    The advantages of the hetero-junction CdS/CuAl2O4 for the photocatalytic eosin degradation are reported. Composite semiconductors are elaborated by co-precipitation of CdS on the spinel CuAl2O4 giving a core-shell structure with a uniform dispersion and intimate contact of the spinel nanoparticles inside the hexagonal CdS. The Mott-Schottky plots ( C -2- V) of both materials show linear behaviors from which flat band potentials are determined. The photoactivity increases with increasing the mass of the sensitizer CdS and the best performance is achieved on CdS/CuAl2O4 (85 %/15 %). The pH has a strong influence on the degradation and the photoactivity peaks at pH 7.78. The dark adsorption eosin is weak (~4 %), hence the change in the eosin concentration is attributed to the photocatalytic process. The degradation follows a zero-order kinetic with a rate constant of 5.2 × 10-8 mol L-1 mn-1 while that of the photolysis is seven times lower (0.75 × 10-8 mol L-1 mn-1).

  9. Human Adenovirus Infection Causes Cellular E3 Ubiquitin Ligase MKRN1 Degradation Involving the Viral Core Protein pVII.

    Science.gov (United States)

    Inturi, Raviteja; Mun, Kwangchol; Singethan, Katrin; Schreiner, Sabrina; Punga, Tanel

    2018-02-01

    Human adenoviruses (HAdVs) are common human pathogens encoding a highly abundant histone-like core protein, VII, which is involved in nuclear delivery and protection of viral DNA as well as in sequestering immune danger signals in infected cells. The molecular details of how protein VII acts as a multifunctional protein have remained to a large extent enigmatic. Here we report the identification of several cellular proteins interacting with the precursor pVII protein. We show that the cellular E3 ubiquitin ligase MKRN1 is a novel precursor pVII-interacting protein in HAdV-C5-infected cells. Surprisingly, the endogenous MKRN1 protein underwent proteasomal degradation during the late phase of HAdV-C5 infection in various human cell lines. MKRN1 protein degradation occurred independently of the HAdV E1B55K and E4orf6 proteins. We provide experimental evidence that the precursor pVII protein binding enhances MKRN1 self-ubiquitination, whereas the processed mature VII protein is deficient in this function. Based on these data, we propose that the pVII protein binding promotes MKRN1 self-ubiquitination, followed by proteasomal degradation of the MKRN1 protein, in HAdV-C5-infected cells. In addition, we show that measles virus and vesicular stomatitis virus infections reduce the MKRN1 protein accumulation in the recipient cells. Taken together, our results expand the functional repertoire of the HAdV-C5 precursor pVII protein in lytic virus infection and highlight MKRN1 as a potential common target during different virus infections. IMPORTANCE Human adenoviruses (HAdVs) are common pathogens causing a wide range of diseases. To achieve pathogenicity, HAdVs have to counteract a variety of host cell antiviral defense systems, which would otherwise hamper virus replication. In this study, we show that the HAdV-C5 histone-like core protein pVII binds to and promotes self-ubiquitination of a cellular E3 ubiquitin ligase named MKRN1. This mutual interaction between the pVII and

  10. 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

  11. 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

  12. Investigation of the coolability of a continuous mass of relocated debris to a water-filled lower plenum. Technical report

    International Nuclear Information System (INIS)

    Rempe, J.L.; Wolf, J.R.; Chavez, S.A.; Condie, K.G.; Hagrman, D.L.; Carmack, W.J.

    1994-09-01

    This report documents work performed to support the development of an analytical and experimental program to investigate the coolability of a continuous mass of debris that relocates to a water-filled lower plenum. The objective of this program is to provide an adequate data base for developing and validating a model to predict the coolability of a continuous mass of debris relocating to a water-filled lower plenum. The model must address higher pressure scenarios, such as the TMI-2 accident, and lower pressure scenarios, which recent calculations indicate are more likely for most operating LWR plants. The model must also address a range of possible debris compositions

  13. Degraded core accidents: review of aerosol behaviour in the containment of a PWR

    International Nuclear Information System (INIS)

    Nichols, A.L.; Walker, B.C.

    1981-09-01

    Low probability-high consequence accidents have become an important issue in reactor safety studies. Such accidents would involve damage to the core and the subsequent release of radioactive fission products into the environment. Aerosols play a major role in the transport and removal of these fission products in the reactor building containment. The aerosol mechanisms, computer modelling codes and experimental studies used to predict aerosol behaviour in the containment of a PWR are reviewed. There are significant uncertainties in the aerosol source terms and specific recommendations have been made for further studies, particularly with respect to code development and high density aerosol-fission product transport within closed systems. (author)

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

    International Nuclear Information System (INIS)

    Oguma, Ritsuo

    1997-01-01

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

  15. SULTAN test facility for large-scale vessel coolability in natural convection at low pressure

    International Nuclear Information System (INIS)

    Rouge, S.

    1997-01-01

    The SULTAN facility (France/CEA/CENG) was designed to study large-scale structure coolability by water in boiling natural convection. The objectives are to measure the main characteristics of two-dimensional, two-phase flow, in order to evaluate the recirculation mass flow in large systems, and the limits of the critical heat flux (CHF) for a wide range of thermo-hydraulic (pressure, 0.1-0.5 MPa; inlet temperature, 50-150 C; mass flow velocity, 5-4400 kg s -1 m -2 ; flux, 100-1000 kW m -2 ) and geometric (gap, 3-15 cm; inclination, 0-90 ) parameters. This paper makes available the experimental data obtained during the first two campaigns (90 , 3 cm; 10 , 15 cm): pressure drop differential pressure (DP) = f(G), CHF limits, local profiles of temperature and void fraction in the gap, visualizations. Other campaigns should confirm these first results, indicating a favourable possibility of the coolability of large surfaces under natural convection. (orig.)

  16. Modeling for evaluation of debris coolability in lower plenum of reactor pressure vessel

    International Nuclear Information System (INIS)

    Maruyama, Yu; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi

    2003-01-01

    Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code. (author)

  17. Hydrocarbon Degradation in Caspian Sea Sediment Cores Subjected to Simulated Petroleum Seepage in a Newly Designed Sediment-Oil-Flow-Through System

    Directory of Open Access Journals (Sweden)

    Tina Treude

    2017-04-01

    Full Text Available The microbial community response to petroleum seepage was investigated in a whole round sediment core (16 cm length collected nearby natural hydrocarbon seepage structures in the Caspian Sea, using a newly developed Sediment-Oil-Flow-Through (SOFT system. Distinct redox zones established and migrated vertically in the core during the 190 days-long simulated petroleum seepage. Methanogenic petroleum degradation was indicated by an increase in methane concentration from 8 μM in an untreated core compared to 2300 μM in the lower sulfate-free zone of the SOFT core at the end of the experiment, accompanied by a respective decrease in the δ13C signal of methane from -33.7 to -49.5‰. The involvement of methanogens in petroleum degradation was further confirmed by methane production in enrichment cultures from SOFT sediment after the addition of hexadecane, methylnapthalene, toluene, and ethylbenzene. Petroleum degradation coupled to sulfate reduction was indicated by the increase of integrated sulfate reduction rates from 2.8 SO42-m-2 day-1 in untreated cores to 5.7 mmol SO42-m-2 day-1 in the SOFT core at the end of the experiment, accompanied by a respective accumulation of sulfide from 30 to 447 μM. Volatile hydrocarbons (C2–C6 n-alkanes passed through the methanogenic zone mostly unchanged and were depleted within the sulfate-reducing zone. The amount of heavier n-alkanes (C10–C38 decreased step-wise toward the top of the sediment core and a preferential degradation of shorter (C30 was seen during the seepage. This study illustrates, to the best of our knowledge, for the first time the development of methanogenic petroleum degradation and the succession of benthic microbial processes during petroleum passage in a whole round sediment core.

  18. Hydrocarbon Degradation in Caspian Sea Sediment Cores Subjected to Simulated Petroleum Seepage in a Newly Designed Sediment-Oil-Flow-Through System.

    Science.gov (United States)

    Mishra, Sonakshi; Wefers, Peggy; Schmidt, Mark; Knittel, Katrin; Krüger, Martin; Stagars, Marion H; Treude, Tina

    2017-01-01

    The microbial community response to petroleum seepage was investigated in a whole round sediment core (16 cm length) collected nearby natural hydrocarbon seepage structures in the Caspian Sea, using a newly developed Sediment-Oil-Flow-Through (SOFT) system. Distinct redox zones established and migrated vertically in the core during the 190 days-long simulated petroleum seepage. Methanogenic petroleum degradation was indicated by an increase in methane concentration from 8 μM in an untreated core compared to 2300 μM in the lower sulfate-free zone of the SOFT core at the end of the experiment, accompanied by a respective decrease in the δ 13 C signal of methane from -33.7 to -49.5‰. The involvement of methanogens in petroleum degradation was further confirmed by methane production in enrichment cultures from SOFT sediment after the addition of hexadecane, methylnapthalene, toluene, and ethylbenzene. Petroleum degradation coupled to sulfate reduction was indicated by the increase of integrated sulfate reduction rates from 2.8 SO 4 2- m -2 day -1 in untreated cores to 5.7 mmol SO 4 2- m -2 day -1 in the SOFT core at the end of the experiment, accompanied by a respective accumulation of sulfide from 30 to 447 μM. Volatile hydrocarbons (C2-C6 n -alkanes) passed through the methanogenic zone mostly unchanged and were depleted within the sulfate-reducing zone. The amount of heavier n -alkanes (C10-C38) decreased step-wise toward the top of the sediment core and a preferential degradation of shorter (C30) was seen during the seepage. This study illustrates, to the best of our knowledge, for the first time the development of methanogenic petroleum degradation and the succession of benthic microbial processes during petroleum passage in a whole round sediment core.

  19. The Analysis of Surrounding Structure Effect on the Core Degradation Progress with COMPASS Code

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Jun Ho; Son, Dong Gun; Kim, Jong Tae; Park, Rae Jun; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In line with the importance of severe accident analysis after Fukushima accident, the development of integrated severe accident code has been launched by the collaboration of three institutes in Korea. KAERI is responsible to develop modules related to the in-vessel phenomena, while other institutes are to the containment and severe accident mitigation facility, respectively. In the first phase, the individual severe accident module has been developed and the construction of integrated analysis code is planned to perform in the second phase. The basic strategy is to extend the design basis analysis codes of SPACE and CAP, which are being validated in Korea for the severe accident analysis. In the first phase, KAERI has targeted to develop the framework of severe accident code, COMPASS (COre Meltdown Progression Accident Simulation Software), covering the severe accident progression in a vessel from a core heat-up to a vessel failure as a stand-alone fashion. In order to analyze the effect of surrounding structure, the melt progression has been compared between the central zone and the most outer zone under the condition of constant radial power peaking factor. Figure 2 and 3 shows the fuel element temperature and the clad mass at the central zone, respectively. Due to the axial power peaking factor, the axial node No.3 has the highest temperature, while the top and bottom nodes have the lowest temperature. When the clad temperature reaches to the Zr melting temperature (2129.15K), the Zr starts to melt. The axial node No.2 reaches to the fuel melting temperature about 5000 sec and the molten fuel relocates to the node No.1, which results to the blockage of flow area in node No.1. The blocked flow area becomes to open about 6100 sec due to the molten ZrO{sub 2} mass relocation to core support plate. Figure 4 and 5 shows the fuel element temperature and the clad mass at the most outer zone, respectively. It is shown that the fuel temperature increase more slowly

  20. Assessment of capability for modeling the core degradation in 2D geometry with ASTEC V2 integral code for VVER type of reactor

    International Nuclear Information System (INIS)

    Dimov, D.

    2011-01-01

    The ASTEC code is progressively becoming the reference European severe accident integral code through in particular the intensification of research activities carried out since 2004. The purpose of this analysis is to assess ASTEC code modelling of main phenomena arising during hypothetical severe accidents and particularly in-vessel degradation in 2D geometry. The investigation covers both early and late phase of degradation of reactor core as well as determination of corium which will enter the reactor cavity. The initial event is station back-out. In order to receive severe accident condition, failure of all active component of emergency core cooling system is apply. The analysis is focus on ICARE module of ASTEC code and particularly on so call MAGMA model. The aim of study is to determine the capability of the integral code to simulate core degradation and to determine the corium composition entering the reactor cavity. (author)

  1. Effect of a blockage length on the coolability during reflood in a 2 × 2 rod bundle with a 90% partially blocked region

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kihwan, E-mail: kihwankim@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Kim, Byung-Jae, E-mail: byoungjae@kaeri.re.kr [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseoung-Gu, Daejeon 34134 (Korea, Republic of); Choi, Hae-Seob, E-mail: hschoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Highlights: • This test was conducted to understand the effect of blockage length on the coolability. • Reflood tests were conducted with blockage simulators for various reflood rates. • The coolability in the downstream of the blockage region is significantly enhanced. - Abstract: If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods owing to the deformed blockage region. The coolability in the blocked region depends on a complex two-phase heat transfer with various thermal hydraulic conditions. In addition, the blockage characteristics, such as the blockage ratio, length, shape, and configurations, are also significant factors affecting the coolability. In the present study, reflood experiments were carried out to understand the effect of the blockage length upon the coolability by varying the reflooding rates. The experiments were performed in electrically heated 2 × 2 rod bundles with blockage simulators having the same blockage ratio but different blockage lengths. The characteristics of quenching and heat transfer were evaluated to investigate the influence of the blockage region on the coolability. The droplet behaviors were also observed by measuring the droplets velocity and size near the blockage region. The coolability in the downstream region of the blockage was significantly enhanced, owing to the reduced flow area of the sub-channel, intensification of turbulence, and the entrained droplets in the blockage region.

  2. Block copolymer micelles with a dual-stimuli-responsive core for fast or slow degradation.

    Science.gov (United States)

    Han, Dehui; Tong, Xia; Zhao, Yue

    2012-02-07

    We report the design and demonstration of a dual-stimuli-responsive block copolymer (BCP) micelle with increased complexity and control. We have synthesized and studied a new amphiphilic ABA-type triblock copolymer whose hydrophobic middle block contains two types of stimuli-sensitive functionalities regularly and repeatedly positioned in the main chain. Using a two-step click chemistry approach, disulfide and o-nitrobenzyle methyl ester groups are inserted into the main chain, which react to reducing agents and light, respectively. With the end blocks being poly(ethylene oxide), micelles formed by this BCP possess a core that can be disintegrated either rapidly via photocleavage of o-nitrobenzyl methyl esters or slowly through cleavage of disulfide groups by a reducing agent in the micellar solution. This feature makes possible either burst release of an encapsulated hydrophobic species from disintegrated micelles by UV light, or slow release by the action of a reducing agent, or release with combined fast-slow rate profiles using the two stimuli.

  3. Coolability of a 3D homogeneous debris bed, experimental and numerical investigations

    International Nuclear Information System (INIS)

    Atkhen, K.; Berthoud, G.

    2001-01-01

    Within the framework of nuclear safety analysis, we present here experimental and numerical results in the field of debris bed coolability. Experimental data are provided by the SILFIDE 3D experimental facility in which the debris bed is heated by induction, at Electricite de France (EDF). Numerical computations are obtained with MC3D-REPO which is a 3-phase and 3D code developed by the Commissariat a l'Energie Atomique (CEA). The uniform debris bed consists of 2 and 3,17 mm diameter steel beads contained in a 50 cm x 60 cm x 10 cm vessel. Water is used as a coolant and can be introduced either by the top or the bottom of the bed at a determined temperature. Due to heterogeneous power distribution within the bed, two definitions for the critical heat flux are proposed: the classical mean value and the local flux (much higher). Even in the first case, the measured dryout heat flux is higher than the Lipinsky 1-D flux. Temperature curve analyses show that the dryout phenomenon is very local, therefore one should be careful about the right flux definition to use. As the injected power is being increased stepwise, steady temperature stages above saturation temperature before dryout can be observed. A discussion is proposed. For some very high values of the induction power, some spheres melted together, leading to a bigger non-porous region. Even if the local temperature went over 1300 C, the bed was still coolable and the critical heat flux value was not impacted. Some parametric studies led to the following conclusions: bottom coolant injection leads to a twice time higher critical flux than by top injection, the influence of the height of the water pool above debris bed is negligible, a sub-cooled liquid injection has no influence on the coolability. Fluidization of surface particles is also discussed. The MC3D-REPO model assumes a thermal equilibrium between the three phases, which gives results in agreement with experiments until dryout occurs. (author)

  4. Selective degradation of model pollutants in the presence of core@shell TiO{sub 2}@SiO{sub 2} photocatalyst

    Energy Technology Data Exchange (ETDEWEB)

    Nadrah, Peter, E-mail: peter.nadrah@zag.si [Slovenian National Building and Civil Engineering Institute, Dimičeva ul. 12, SI-1000 Ljubljana (Slovenia); Gaberšček, Miran [National Institute of Chemistry, Hajdrihova ul. 19, SI-1000 Ljubljana (Slovenia); Sever Škapin, Andrijana [Slovenian National Building and Civil Engineering Institute, Dimičeva ul. 12, SI-1000 Ljubljana (Slovenia)

    2017-05-31

    Highlights: • TiO{sub 2} encapsulated in mesoporous silica exhibits selective photocatalytic degradation of low-molecular-weight molecules. • Core@shell photocatalyst degrades rhodamine B in presence of fivefold mass concentration of starch, while pure TiO{sub 2} does not. • Potential use for removing water pollutants, while retaining non-harmful and beneficial macromolecules. - Abstract: Photocatalytic TiO{sub 2} degrades organic matter unselectively. However, in certain applications, such as degradation of pollutants, selectivity towards pollutants is beneficial. We synthesized core@shell TiO{sub 2}@SiO{sub 2} nanoparticles with photocatalytic activity featuring a significantly faster preferential degradation of model pollutant (rhodamine B) in presence of abundant concentration of natural organic matter compared to pure TiO{sub 2} (P25). The material’s photocatalytic activity was tested in aqueous medium. The selectivity of prepared effect of core@shell materials is explained based on transmission electron microscopy, nitrogen adsorption, X-ray powder diffraction and zeta potential measurements.

  5. A comparison of core degradation phenomena in the CORA, QUENCH, Phébus SFD and Phébus FP experiments

    Energy Technology Data Exchange (ETDEWEB)

    Haste, T., E-mail: tim.haste@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Steinbrück, M., E-mail: martin.steinbrueck@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Barrachin, M., E-mail: marc.barrachin@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Luze, O. de, E-mail: olivier.de-luze@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN, BP 3, F-13115 St. Paul-lez-Durance Cedex (France); Grosse, M., E-mail: mirco.grosse@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Stuckert, J., E-mail: juri.stuckert@kit.edu [Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)

    2015-03-15

    Highlights: • The results of the experiments CORA, QUENCH and Phébus SFD/FP are summarised. • All phenomena expected up to melt movement to the lower head are shown consistently. • Separate-effect tests performed at KIT and IRSN aid improve their modelling. • Data from the integral tests help independent validation of new and improved models. • The improved codes will help reduce uncertainties in safety-critical areas for core degradation. - Abstract: Over the past 20 years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP – fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH – electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that occur under severe accident conditions, relevant to such safety issues as in-vessel retention of the core, recovery of the core by water reflood, hydrogen generation and fission product release. These data form an important basis for development and validation of severe accident analysis codes such as ASTEC (IRSN/GRS, EC) and MELCOR (USNRC/SNL, USA) that are used to assess the safety of current and future reactor designs, so helping to reduce the uncertainty associated with such code predictions. Following the recent end of the Phébus FP project, it is appropriate now to compare the core degradation phenomena observed in these four major experimental series, indicating the main conclusions that have been drawn. This covers subjects such as early phase degradation up to loss of rod-like geometry (all the series), late phase degradation and the link between fission product release and core degradation (Phébus FP), oxidation phenomena (all the series), reflood behaviour (CORA and QUENCH), as well as particular topics such as the effects of control rod material and fuel burn-up on core degradation. It also outlines the separate-effects experiments performed to

  6. Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

    International Nuclear Information System (INIS)

    Ginsberg, T.; Chen, J.C.

    1984-01-01

    Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

  7. A review of dryout heat fluxes and coolability of particle beds. APRI 4, Stage 2 Report

    International Nuclear Information System (INIS)

    Lindholm, Ilona

    2002-04-01

    were studied. Significant amount of data with prototypic material tests exists. All of the tests show significant fragmentation in case of deep subcooled pool. An additional observation is that no energetic melt coolant interaction (steam explosion) has been reported for prototypic materials. A set of most relevant data for reactor applications have been chosen. Based on this, a general particle size distribution has been constructed. The average particle size obtained by this way was about 3.5 mm. Information from fragmentation and dryout tests and the Lipinski 0-D correlation have been utilised to assess the debris bed coolability for the Olkiluoto severe accident scenario. The calculation shows that for well-mixed beds with 3.5 mm particles the dryout heat flux would be close to 1 MW/m 2 , well above the estimated heat flux due to decay heat. Stratification of finer particles on top of the bed due to e.g. a steam explosion would reduce the dryout heat flux to 50-200 kW/m 2 . This would be below heat fluxes produced by decay heat in Nordic BWRs. The key uncertainty considering particle bed coolability is due to the particle size distribution and stratification. If the possibility of a thick fine particle layer on top of the bed can be ruled out, the particulate debris bed in Nordic BWRs will be coolable. A rough estimate of melt pool coolability in Nordic BWRs has also been conducted. The MACE and COTELS experimental data have been summarised. Based on the data, the melt pools in the pedestal are slowly coolable. The concrete erosion does not threaten the containment failure margins, except maybe at Forsmark 1 and 2 units. Release of non-condensable gases during MCCI may cause an earlier start of filtered venting in Olkiluoto, Forsmark and Oskarshamn 3 plants

  8. Improvement and evaluation of debris coolability analysis module in severe accident analysis code SAMPSON using LIVE experiment

    International Nuclear Information System (INIS)

    Wei, Hongyang; Erkan, Nejdet; Okamoto, Koji; Gaus-Liu, Xiaoyang; Miassoedov, Alexei

    2017-01-01

    Highlights: • Debris coolability analysis module in SAMPSON is validated. • Model for heat transfer between melt pool and pressure vessel wall is modified. • Modified debris coolability analysis module is found to give reasonable results. - Abstract: The purpose of this work is to validate the debris coolability analysis (DCA) module in the severe accident analysis code SAMPSON by simulating the first steady stage of the LIVE-L4 test. The DCA module is used for debris cooling in the lower plenum and for predicting the safety margin of present reactor vessels during a severe accident. In the DCA module, the spreading and cooling of molten debris, gap cooling, heating of a three-dimensional reactor vessel, and natural convection heat transfer are all considered. The LIVE experiment is designed to investigate the formation and stability of melt pools in a reactor pressure vessel (RPV). By comparing the simulation results and experimental data in terms of the average melt pool temperature and the heat flux along the vessel wall, a bug is found in the code and the model for the heat transfer between the melt pool and RPV wall is modified. Based on the Asfia–Dhir and Jahn–Reineke correlations, the modified version of the DCA module is found to give reasonable results for the average melt pool temperature, crust thickness in the steady state, and crust growth rate.

  9. Study on effective particle diameters and coolability of particulate beds packed with irregular multi-size particles

    Energy Technology Data Exchange (ETDEWEB)

    Thakre, S.; Ma, W.; Kudinov, P.; Bechta, S. [Royal Institute of Technology, KTH. Div. of Nuclear Power Safety, Stockholm (Sweden)

    2013-08-15

    One of the key questions in severe accident research is the coolability of the debris bed, i.e., whether decay heat can be completely removed by the coolant flow into the debris bed. Extensive experimental and analytical work has been done to substantiate the coolability research. Most of the available experimental data is related to the beds packed with single size (mostly spherical) particles, and less data is available for multi-size/irregular-shape particles. There are several analytical models available, which rely on the mean particle diameter and porosity of the bed in their predictions. Two different types of particles were used to investigate coolability of particulate beds at VTT, Finland. The first type is irregular-shape Aluminum Oxide gravel particles whose sizes vary from 0.25 mm to 10 mm, which were employed in the STYX experiment programme (2001-2008). The second type is spherical beads of Zirconium silicate whose sizes vary between 0.8 mm to 1 mm, which were used in the COOLOCE tests (Takasuo et al., 2012) to study the effect of multi-dimensional flooding on coolability. In the present work, the two types of particles are used in the POMECO-FL and POMECO-HT test facility to obtain their effective particle diameters and dryout heat flux of the beds, respectively. The main idea is to check how the heaters' orientations (vertical in COOLOCE vs. horizontal in POMECO-HT) and diameters (6 mm in COOLOCE vs. 3 mm in POMECO-HT) affect the coolability (dryout heat flux) of the test beds. The tests carried out on the POMECO-FL facility using a bed packed with aluminum oxide gravel particles show the effective particle diameter of the gravel particles is 0.65 mm, by which the frictional pressure gradient can be predicted by the Ergun equation. After the water superficial velocity is higher than 0.0025 m/s, the pressure gradient is underestimated. The effective particle diameter of the zirconium particles is found as 0.8 mm. The dryout heat flux is measured on

  10. Study on effective particle diameters and coolability of particulate beds packed with irregular multi-size particles

    International Nuclear Information System (INIS)

    Thakre, S.; Ma, W.; Kudinov, P.; Bechta, S.

    2013-08-01

    One of the key questions in severe accident research is the coolability of the debris bed, i.e., whether decay heat can be completely removed by the coolant flow into the debris bed. Extensive experimental and analytical work has been done to substantiate the coolability research. Most of the available experimental data is related to the beds packed with single size (mostly spherical) particles, and less data is available for multi-size/irregular-shape particles. There are several analytical models available, which rely on the mean particle diameter and porosity of the bed in their predictions. Two different types of particles were used to investigate coolability of particulate beds at VTT, Finland. The first type is irregular-shape Aluminum Oxide gravel particles whose sizes vary from 0.25 mm to 10 mm, which were employed in the STYX experiment programme (2001-2008). The second type is spherical beads of Zirconium silicate whose sizes vary between 0.8 mm to 1 mm, which were used in the COOLOCE tests (Takasuo et al., 2012) to study the effect of multi-dimensional flooding on coolability. In the present work, the two types of particles are used in the POMECO-FL and POMECO-HT test facility to obtain their effective particle diameters and dryout heat flux of the beds, respectively. The main idea is to check how the heaters' orientations (vertical in COOLOCE vs. horizontal in POMECO-HT) and diameters (6 mm in COOLOCE vs. 3 mm in POMECO-HT) affect the coolability (dryout heat flux) of the test beds. The tests carried out on the POMECO-FL facility using a bed packed with aluminum oxide gravel particles show the effective particle diameter of the gravel particles is 0.65 mm, by which the frictional pressure gradient can be predicted by the Ergun equation. After the water superficial velocity is higher than 0.0025 m/s, the pressure gradient is underestimated. The effective particle diameter of the zirconium particles is found as 0.8 mm. The dryout heat flux is measured on

  11. Modeling of reflood of severely damaged reactor core

    International Nuclear Information System (INIS)

    Bachrata, A.

    2012-01-01

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

  12. 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.

  13. Efficient photocatalytic degradation of malachite green dye under visible irradiation by water soluble ZnS:Mn/ZnS core/shell nanoparticles

    Science.gov (United States)

    Khaparde, Rohini A.; Acharya, Smita A.

    2018-05-01

    ZnS:Mn/ ZnS core/shell nanoparticles was prepared by two step synthesis method. In first step, oleic acid - coated Mn doped ZnS core nanoparticles were prepared which were charged through ligand exchange. Shell of ZnS NPs was finally deposited upon the surface of charged Mn doped ZnS core. Scanning electron microscopy (SEM) image exhibit morphological confirmation of ZnS:Mn/ZnS core/shell. As Nano ZnS are the most suitable candidates for photocatalyst that extensively involved in degradation and complete mineralization of various toxic organic pollutants owing to its high efficiency, strong oxidizing power, non-toxicity, high photochemical and biological stability, corrosive resistance and low cost. Photodegradation of malachite green is systematically investigated by adding different molar proportional of ZnS:Mn/ZnS core/shell in the dye. The rate of de-coloration of dye is detected by UV-VIS absorption spectroscopy. Efficient detoriation in the colour of dye is attributed to the core /shell morphology of the particles.

  14. Enhancement of Fenton processes at initial circumneutral pH for the degradation of norfloxacin with Fe@Fe2O3 core-shell nanomaterials.

    Science.gov (United States)

    Liu, Jingyi; Hu, Wenyong; Sun, Maogui; Xiong, Ouyang; Yu, Haibin; Feng, Haopeng; Wu, Xuan; Tang, Lin; Zhou, Yaoyu

    2018-06-13

    The degradation of norfloxacin by Fenton reagent with core-shell Fe@Fe 2 O 3 nanomaterials was studied under neutral conditions in a closed batch system. Norfloxacin was significantly degraded (90%) in the Fenton system with Fe@Fe 2 O 3 in 30 min at the initial pH 7.0, but slightly degraded in Fenton system without Fe@Fe 2 O 3 under the same experimental conditions. The intermediate products were investigated by gas chromatography-mass spectrometry, and the possible Fenton oxidation pathway of norfloxacin in the presence of Fe@Fe 2 O 3 nanowires was proposed. Electron spin resonance spectroscopy was used to identify and characterize the free radicals generated, and the mechanism for norfloxacin degradation was also revealed. Finally, the reusability and the stability of Fe@Fe 2 O 3 nanomaterials were studied using x-ray diffraction and scanning electron microscope, which indicated that Fe@Fe 2 O 3 is a stable catalyst and can be used repetitively in environmental pollution control.

  15. Analyzing different HPCI operation modes simulated with ATHLET-CD regarding possible core degradation phenomena in Fukushima-Daiichi unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Bratfisch, Christoph; Koch, Marco K. [Ruhr-Univ. Bochum (Germany). Reactor Simulation and Safety Group

    2017-02-15

    For extented application and analyses of the severe accident code ATHLET-CD, the course of the invessel accident in Unit 3 of Fukushima-Daiichi is simulated in the frame of the research project SUBA as a part of the BMBF sponsored collaborative project WASA-BOSS (Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Stoerfallmassnahmen). Investigations, carried out by TEPCO, had shown that the High-Pressure Coolant Injection system (HPCI) might have stopped earlier than expected. A parameter variation was performed to analyze the impact of the tripped HPCI injection regarding the thermohydraulic behaviour as well as the core degradation phenomena.

  16. Application of Ni-Oxide@TiO2 Core-Shell Structures to Photocatalytic Mixed Dye Degradation, CO Oxidation, and Supercapacitors

    Directory of Open Access Journals (Sweden)

    Seungwon Lee

    2016-12-01

    Full Text Available Performing diverse application tests on synthesized metal oxides is critical for identifying suitable application areas based on the material performances. In the present study, Ni-oxide@TiO2 core-shell materials were synthesized and applied to photocatalytic mixed dye (methyl orange + rhodamine + methylene blue degradation under ultraviolet (UV and visible lights, CO oxidation, and supercapacitors. Their physicochemical properties were examined by field-emission scanning electron microscopy, X-ray diffraction analysis, Fourier-transform infrared spectroscopy, and UV-visible absorption spectroscopy. It was shown that their performances were highly dependent on the morphology, thermal treatment procedure, and TiO2 overlayer coating.

  17. Preparation of ZnS@In2S3 Core@shell Composite for Enhanced Photocatalytic Degradation of Gaseous o-Dichlorobenzene under Visible Light.

    Science.gov (United States)

    Liu, Baojun; Hu, Xia; Li, Xinyong; Li, Ying; Chen, Chang; Lam, Kwok-Ho

    2017-11-27

    In this study, novel ZnS@In 2 S 3 core@shell hollow nanospheres were fabricated by a facile refluxing method for the first time, and the formation mechanism of hollow structure with interior architecture was discussed based on ion-exchange Ostwald ripening. As the photocatalytic material for degradation of gaseous o-Dichlorobenzene (o-DCB), the as-synthesized core@shell hollow nanospheres were found to show significantly enhanced catalytic performance for effective separation of photo-generated charges. Moreover, the mechanisms of enhanced activity were elucidated by band alignment and unique configuration. Such photocatalyst would meet the demands for the control of persistent organic pollutant (POPs) in the atmospheric environment.

  18. Analysis of the thermal hydraulics and core degradation behavior in the PHEBUS-FPT1 test train with impact/SAMPSON code

    International Nuclear Information System (INIS)

    Terada, Masafumi; Ikeda, Takashi; Nakahara, Katsuhiko; Shirakawa, Noriyuki; Horie, Hideki; Katsuragi, Kazuyuki; Yamagishi, Makoto; Ito, Takahiro

    2003-01-01

    As one of the verification studies of SAMPSON code, PHEBUS-FPT1, which is authorized as the International Standard Problem-46, was analyzed about the in-core phenomena with four modules, the molten core relocation analysis (MCRA) module, the fuel rod heat up analysis (FRHA) module, the fission product release analysis (FPRA) module, and the analysis control module (ACM) of SAMPSON. This paper describes the analysis of thermal hydraulics and core degradation behavior in the test train. Two-dimensional version of MCRA models the whole structure of the test train in the cylindrical system, including the fuel bundle and the shroud. FRHA models eighteen irradiated fuel rods, two fresh fuel rods, and one control rod in the center of the bundle. FRHA evaluates the transient behavior of fuel rods and releases failed fuel components to MCRA. MCRA evaluates the fluid dynamics of steam and debris considering the thermal and fluid mechanical interaction between them, and at the same time the thermal interaction between gas/debris and shroud material. By the phase change model of MCRA, molten debris forms debris pool and a part of them possibly freezes on fuel rods or shroud surface, then forms crust. This combination of modules of SAMPSON was proved to be capable for modeling the PHEBUS-FPT1 in-core phenomena sufficiently. The analysis has shown sufficient agreement with test results regarding to steam flow rates at the outlet, reproducing its reduction due to hydrogen generation, steam and shroud temperature, and debris relocation behavior. (author)

  19. The TMI-2 core relocation: Heat transfer and mechanism

    International Nuclear Information System (INIS)

    Epstein, M.; Fauske, H.K.

    1987-07-01

    It is postulated that the collapse of the upper debris bed was the main cause of core failure and core material relocation during the TMI-2 accident. It is shown that this mechanism of core relocation can account for the timescale(s) and energy transfer rate inferred from plant instrumentation. Additional analysis suggests that the water in the lower half of the reactor vessel was subcooled at the onset of relocation, as subcooling serves to explain the final coolable configuration at the bottom of the TMI vessel

  20. Status of the corquench model for calculation of ex-vessel corium coolability by an overlying water layer

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.

    2000-01-01

    The results of melt attack and coolability experiment (MACE) tests have identified several heat transfer mechanisms which could potentially lead to long term corium coolability. Based on physical observations from these tests, an integrated model of corium quenching (CORQUENCH) behavior is being developed. Aside from modeling of the primary physical processes observed in the tests, considerable effort has also been devoted to modeling of test occurrences which deviate from the behavior expected at reactor scale. In this manner, extrapolation of the models validated against the test data to the reactor case can be done with increased confidence. The integrated model currently addresses early bulk cooling and incipient crust formation heat transfer phases, as well as a follow-on water ingression phase which leads to development of a sustained quench front progressing downwards through the debris. In terms of experiment distortions, the model is also able to mechanistically calculate crust anchoring to the test section sidewalls, as well as the subsequent melt/crust separation phase which arises due to concrete densification upon melting. In this paper, the status of the model development and validation activities are described. In addition, representative calculations for PWR plant conditions are provided in order to illustrate the potential benefits of overlying water on mitigation of the accident sequence. (orig.)

  1. COOLOCE debris bed experiments and simulations investigating the coolability of cylindrical beds with different materials and flow modes

    Energy Technology Data Exchange (ETDEWEB)

    Takasuo, E.; Kinnunen, T.; Holmstroem, S.; Lehtikuusi, T. [VTT Technical Research Centre of Finland (Finland)

    2013-07-15

    The COOLOCE experiments aim at investigating the coolability of debris beds of different geometries, flow modes and materials. A debris bed may be formed of solidified corium as a result of a severe accident in a nuclear power reactor. The COOLOCE-8 test series consisted of experiments with a top-flooded test bed with irregular gravel as the simulant material. The objective was to produce comparison data useful in estimating the effects of different particle materials and the possible effect of the test arrangement on the results. It was found that the dryout heat flux (DHF) measured for the gravel was lower compared to previous experiments with spherical beads, and somewhat lower compared to the early STYX experiments. The difference between the beads and gravel is at least partially explained by the smaller average size of the gravel particles. The COOLOCE-9 test series included scoping experiments examining the effect of subcooling of the water pool in which the debris bed is immersed. The experiments with initially subcooled pool suggest that the subcooling may increase DHF and increase coolability. The aim of the COOLOCE-10 experiments was to investigate the effect of lateral flooding on the DHF a cylindrical test bed. The top of the test cylinder and its sidewall were open to water infiltration. It was found that the DHF is increased compared to a top-flooded cylinder by more than 50%. This suggests that coolability is notably improved. 2D simulations of the top-flooded test beds have been run with the MEWA code. Prior to the simulations, the effective particle diameter for the spherical beads and the irregular gravel was estimated by single-phase pressure loss measurements performed at KTH in Sweden. Parameter variations were done for particle size and porosity used as input in the models. It was found that with the measured effective particle diameter and porosity, the simulation models predict DHF with a relatively good accuracy in the case of spherical

  2. Combination chemotherapy using core-shell nanoparticles through the self-assembly of HPMA-based copolymers and degradable polyester

    Czech Academy of Sciences Publication Activity Database

    Jäger, Eliezer; Jäger, Alessandro; Chytil, Petr; Etrych, Tomáš; Říhová, Blanka; Giacomelli, F. C.; Štěpánek, Petr; Ulbrich, Karel

    2013-01-01

    Roč. 165, č. 2 (2013), s. 153-161 ISSN 0168-3659 R&D Projects: GA AV ČR IAAX00500803; GA ČR GA202/09/2078; GA ČR GPP207/11/P551 Institutional research plan: CEZ:AV0Z40500505; CEZ:AV0Z50200510 Institutional support: RVO:61389013 ; RVO:61388971 Keywords : combination therapy * polymeric core-shell nanoparticles * docetaxel Subject RIV: CD - Macromolecular Chemistry; EC - Immunology (MBU-M) Impact factor: 7.261, year: 2013

  3. Numerical simulation of the insulation material transport to a PWR core under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    Höhne, Thomas; Grahn, Alexander; Kliem, Sören; Rohde, Ulrich; Weiss, Frank-Peter

    2013-01-01

    Highlights: ► Detailed results of a numerical simulation of the insulation material transport to a PWR core are shown. ► The spacer grid is modeled as a strainer which completely retains the insulation material carried by coolant. ► The CFD calculations showed that the fibers at the upper spacer grid plane are not uniformly distributed. ► Furthermore the pressure loss does not exceed a critical limit. ► The PWR core coolablity can be guaranteed all the time during the transient. -- Abstract: In 1992, strainers on the suction side of the ECCS pumps in Barsebäck NPP Unit 2 became partially clogged with mineral wool because after a safety valve opened the steam impinged on thermally insulated equipment and released mineral wool. This event pointed out that strainer clogging is an issue in the course of a loss-of-coolant accident. Modifications of the insulation material, the strainer area and mesh size were carried out in most of the German NPPs. Moreover, back flushing procedures to remove the mineral wool from the strainers and differential pressure measurements were implemented to assure the performance of emergency core cooling during the containment sump recirculation mode. Nevertheless, it cannot be completely ruled out, that a limited amount of small fractions of the insulation material is transported into the RPV. During a postulated cold leg LOCA with hot leg ECC injection, the fibers enter the upper plenum and can accumulate at the fuel element spacer grids, preferably at the uppermost grid level. This effect might affect the ECC flow into the core and could result in degradation of core cooling. It was the aim of the numerical simulations presented to study where and how many mineral wool fibers are deposited at the upper spacer grid. The 3D, time dependent, multi-phase flow problem was modeled applying the CFD code ANSYS CFX. The CFD calculation does not yet include steam production in the core and also does not include re-suspension of the

  4. One-step synthesis, toxicity assessment and degradation in tumoral pH environment of SiO2@Ag core/shell nanoparticles

    Science.gov (United States)

    De Matteis, Valeria; Rizzello, Loris; Di Bello, Maria Pia; Rinaldi, Rosaria

    2017-06-01

    The unique physicochemical properties of SiO2@Ag core/shell nanoparticles make them a promising tool in nanomedicine, where they are used as nanocarriers for several biomedical applications, including (but not restricted to) cancer treatment. However, a comprehensive estimation of their potential toxicity, as well as their degradation in the tumor microenvironment, has not been extensively addressed yet. We investigated in vitro the viability, the reactive oxygen species (ROS) production, the DNA damage level, and the nanoparticle uptake on HeLa cells, used as model cancer cells. In addition, we studied the NPs degradation profile at pH 6.5, to mimic the tumor microenvironment, and at the neutral and physiological (pH 7-7.4). Our experiments demonstrate that the silver shell dissolution is promoted under acidic conditions, which could be related to cell death induction. Our evidences demonstrate that SiO2@Ag nanoparticles possess the ability of combining an effective cancer cell treatment (through local silver ions release) together with a possible controlled release of bioactive compounds encapsulated in the silica as future application.

  5. LWR fuel cladding deformation in a LOCA and its interaction with the emergency core cooling

    International Nuclear Information System (INIS)

    Erbacher, F.J.

    1982-01-01

    The paper summarizes research results of out-of-pile burst tests, in-pile bursts tests, out-of-pile flooding tests and modeling work on fuel behavior in a LOCA performed at KfK: The dominant phenomena of the cladding deformation and failure have been clarified by experiments and can be modeled by computer codes. The burst and flooding tests performed up to now suggest that the coolability of the core under LOCA conditions can be maintained. (orig.) [de

  6. Quench cooling of superheated debris beds in containment during LWR core meltdown accidents

    International Nuclear Information System (INIS)

    Ginsberg, T.; Chen, J.C.

    1984-01-01

    Light water reactor core meltdown accident sequence studies suggest that superheated debris beds may settle on the concrete floor beneath the reactor vessel. A model for the heat transfer processes during quench (removal of stored energy from initial temperature to saturation temperature) of superheated debris beds cooled by an overlying pool of water has been presented in a prior paper. This paper discusses the coolability of decay-heated debris beds from the standpoint of their transient quench characteristics. It is shown that even though a debris bed configuration may be coolable from the point of view of steady-state decay heat removal, the quench behavior from an initially elevated temperature may lead to bed melting prior to quench of the debris

  7. 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)

  8. Ex-vessel debris coolability test during severe accident (COTELS project)

    International Nuclear Information System (INIS)

    Ogasawara, H.

    1998-01-01

    The objectives of the COTELS project are for severe accident management, to investigate phenomena of ex-vessel fuel-coolant interactions after reactor pressure vessel (RPV) failure and to investigate molten core-concrete interaction when coolant is injected onto molten debris. The project has being cooperated with the National Nuclear Center in the Republic of Kazakstan from 1994 to 1997 under the sponsorship of the Ministry of International Trade and Industry of Japan. Total programs are composed with the following tests. (1) Test 01 was meant to observe flow mode of falling debris. (2) Test A was meant to investigate phenomena of fuel-coolant interactions when molten debris falls into a coolant pool. (3) Test B/C investigated fuel coolant interactions and molten core-concrete interaction when coolant is injected onto debris. Detail data evaluation is underway. The following results were thus for obtained: (1) It was confirmed in Test 01 series that about 60 kg of UO 2 mixture was completely melted and fallen as a continuous jet. (2) No energetic fuel-coolant interaction was observed both in Test A and B series. (3) Debris in which decay heat was simulated was cooled by water injection in Test C series

  9. Numerical models for the analysis of thermal behavior and coolability of a particulate debris bed in reactor lower head

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Sang Baik; Kim, Byung Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-04-01

    This report provides three distinctive, but closely related numerical models developed for the analysis of thermal behavior and coolability of a particulate debris bed that is may be formed inside the reactor lower head during severe accident late phases. The first numerical module presented in the report, MELTPRO-DRY, is used to analyze numerically heat-up and melting process of the dry particle bed, downward- and sideward-relocation of the liquid melt under gravity force and capillary force acting among porous particles, and solidification of the liquid melt relocated into colder region. The second module, MELTPROG-WET, is used to simulate numerically the cooling process of the particulate debris bed under the existence of water, which is subjected to two types of numerical models. The first type of WET module utilizes distinctive models that parametrically simulate the water cooling process, that is, quenching region, dryout region, and transition region. The choice of each parametric model depends on temperature gradient between the cooling water and the debris particles. The second type of WET module utilizes two-phase flow model that mechanically simulates the cooling process of the debris bed. For a consistent simulation from the water cooling to the dryout debris bed, on the other hand, the aforementioned two modules, MELTPROG-DRY and MELTPROG-WET, were integrated into a single computer program DBCOOL. Each of computational models was verified through limited applications to a heat-generating particulate bed contained in the rectangular cavity. 22 refs., 5 figs., 2 tabs. (Author)

  10. Second OECD (NEA) CSNI specialist meeting on molten core debris-concrete interactions

    International Nuclear Information System (INIS)

    Alsmeyer, H.

    1992-11-01

    The 37 contributions concentrated on two main topics. The first topic is the 'classical' core debris-concrete interaction, both experimental and theoretical. Integral effects and separate effects were addressed in thermal hydraulics and heat transfer, material interaction, and aerosol release during concrete erosion, with some applications to prototypical nuclear power plants. The second topic is the possibility of controlling and ending the erosion of the concrete by spreading of the core melt, and/or achieving coolability by the addition of water. (orig./HP) [de

  11. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    Bandini, G.; Caira, M.; Naviglio, A.; Sorabella, L.

    1995-01-01

    The MARS nuclear plant is equipped with a 600 MWth PWR type nuclear steam supply system, with completely innovative engineered core safeguards. The most relevant innovative safety system of this plant is its Emergency Core Cooling System, which is completely passive (with only one non static component). The Emergency Core Cooling System (ECCS) of the MARS reactor is natural-circulation, passive-type, and its intervention follows a core flow decrease, whatever was the cause. The operation of the system is based on a cascade of three fluid systems, functionally interfacing through heat exchangers; the first fluid system is connected to the reactor vessel and the last one includes an atmospheric-pressure condenser, cooled by external air. The infinite thermal capacity of the final heat sink provides the system an unlimited autonomy. The capability and operability of the system are based on its integrity and on the integrity of the primary coolant boundary (both of them are permanently enclosed in a pressurized containment; 100% redundancy is also foreseen) and on the operation of only one non static component (a check valve), with 400% redundancy. In the paper, all main thermal hydraulic transients occurring as a consequence of postulated accidents are analysed, to verify the capability of the passive-type ECCS to intervene always in time, without causing undue conditions of reduced coolability of the core (DNB, etc.), and to verify its capability to guarantee a long-term (indefinite) coolability of the core without the need of any external intervention. (author)

  12. Green synthesis of the reduced graphene oxide–CuI quasi-shell–core nanocomposite: A highly efficient and stable solar-light-induced catalyst for organic dye degradation in water

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jiha; Reddy, D. Amaranatha; Islam, M. Jahurul [Department of Chemistry and Chemical Institute for Functional Materials, Pusan National University, Busan 609-735 (Korea, Republic of); Seo, Bora [Department of Chemistry, Ulsan National Institute of Science and Technology (UNIST), Ulsan 689-798 (Korea, Republic of); Joo, Sang Hoon [Department of Chemistry, Ulsan National Institute of Science and Technology (UNIST), Ulsan 689-798 (Korea, Republic of); School of Energy and Chemical Engineering, Ulsan National Institute of Science and Technology (UNIST), Ulsan 689-798 (Korea, Republic of); Kim, Tae Kyu, E-mail: tkkim@pusan.ac.kr [Department of Chemistry and Chemical Institute for Functional Materials, Pusan National University, Busan 609-735 (Korea, Republic of)

    2015-12-15

    Graphical abstract: - Highlights: • Green synthesis of RGO–CuI quasi-shell–core nanocomposites without any surfactant. • Promising candidates as solar light active photocatalyst for dye degradation. • Significant improvement of the photocatalytic activity in RGO wrapped composites. • The best photocatalytic activity to RhB has been attained for CuI–RGO (2 mg mL{sup −1}). - Abstract: Surfactant-free, reduced graphene oxide (RGO)–CuI quasi-shell−core nanocomposites were successfully synthesized using ultra-sonication assisted chemical method at room temperature. The morphologies, structures and optical properties of the CuI and CuI–RGO nanocomposites were characterized by transmission electron microscopy (TEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Fourier-transformed infrared spectroscopy (FTIR), UV–visible absorption spectroscopy, and photoluminescence (PL) spectroscopy. Morphological and structural analyses indicated that the CuI–RGO core–shell nanocomposites comprise single-crystalline face-centered cubic phase CuI nanostructures, coated with a thin RGO quasi-shell. Photocatalysis experiments revealed that the as-synthesized CuI–RGO nanocomposites exhibit remarkably enhanced photocatalytic activities and stabilities for photo degradation of Rhodamine-B (RhB) organic dye under simulated solar light irradiation. The photo degradation ability is strongly affected by the concentration of RGO in the nanocomposites; the highest photodegradation rate was obtained at a graphene loading content of 2 mg mL{sup −1} nanocomposite. The remarkable photocatalytic performance of the CuI–RGO nanocomposites mainly originates from their unique adsorption and electron-accepting and electron-transporting properties of RGO. The present work provides a novel green synthetic route to producing CuI–RGO nanocomposites without toxic solvents or reducing agents, thereby providing highly efficient and stable solar light

  13. Green synthesis of the reduced graphene oxide–CuI quasi-shell–core nanocomposite: A highly efficient and stable solar-light-induced catalyst for organic dye degradation in water

    International Nuclear Information System (INIS)

    Choi, Jiha; Reddy, D. Amaranatha; Islam, M. Jahurul; Seo, Bora; Joo, Sang Hoon; Kim, Tae Kyu

    2015-01-01

    Graphical abstract: - Highlights: • Green synthesis of RGO–CuI quasi-shell–core nanocomposites without any surfactant. • Promising candidates as solar light active photocatalyst for dye degradation. • Significant improvement of the photocatalytic activity in RGO wrapped composites. • The best photocatalytic activity to RhB has been attained for CuI–RGO (2 mg mL −1 ). - Abstract: Surfactant-free, reduced graphene oxide (RGO)–CuI quasi-shell−core nanocomposites were successfully synthesized using ultra-sonication assisted chemical method at room temperature. The morphologies, structures and optical properties of the CuI and CuI–RGO nanocomposites were characterized by transmission electron microscopy (TEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Fourier-transformed infrared spectroscopy (FTIR), UV–visible absorption spectroscopy, and photoluminescence (PL) spectroscopy. Morphological and structural analyses indicated that the CuI–RGO core–shell nanocomposites comprise single-crystalline face-centered cubic phase CuI nanostructures, coated with a thin RGO quasi-shell. Photocatalysis experiments revealed that the as-synthesized CuI–RGO nanocomposites exhibit remarkably enhanced photocatalytic activities and stabilities for photo degradation of Rhodamine-B (RhB) organic dye under simulated solar light irradiation. The photo degradation ability is strongly affected by the concentration of RGO in the nanocomposites; the highest photodegradation rate was obtained at a graphene loading content of 2 mg mL −1 nanocomposite. The remarkable photocatalytic performance of the CuI–RGO nanocomposites mainly originates from their unique adsorption and electron-accepting and electron-transporting properties of RGO. The present work provides a novel green synthetic route to producing CuI–RGO nanocomposites without toxic solvents or reducing agents, thereby providing highly efficient and stable solar light-induced RGO

  14. Analysis of heat transfer mechanism on in-vessel corium coolability in severe accidents

    International Nuclear Information System (INIS)

    Park, Rae Joon; Jeong, Ji Whan; Kim, Sang Baik; Kang, Kyung Ho; Kim, Jong Whan

    1998-04-01

    When the molten core material relocates to the lower plenum of the reactor vessel, the cooling process of corium and the related heat transfer mechanism have been analyzed. The critical heat flux in gap (CHFG) test is being performed as a part of simulation of naturally arrested thermal attack in (SONATA-IV) project and the state of art on CHF has been reviewed. A series of complex heat transfer mechanism of molten pool formation, natural convection in the molten pool, solidification and remelting of the corium, conduction in the solidified crust, and boiling heat transfer to surroundings can be occurred in the lower plenum. Many studies are needed to investigate the complex heat transfer mechanism in the lower plenum, because these phenomena have not been clearly understand until now. The SONATA-IV/CHFG experiments are being carried out to develop CHF correlation in a hemispherical gap, which is the upper limit of heat transfer. There is no experimental or analytical CHF correlation applicable to a hemispherical gap. So lots of analytical and experimental correlations developed using the similar experimental condition were gathered and compared with each other. According to the experimental work that was carried out with pool boiling condition, CHF in a parallel gap was reduced by 1/30 compared with the value measured without gap. A basic form of a CHF correlation has been developed to correlate measurements that will be made in the SONATA-IV/CHFG experiments. That correlation is based on the fact that the CHF in a hemispherical gap is enhanced by CCFL and a Kutateladze type CCFL correlation develops CCFL date will in geometry like this. The experimental facility consists of a heater, a pressure vessel, a heat exchanger and lots of sensors. The heater capacity is 40 kw and the maximum heat flux at the surface is 100 kw/m 2 . The experiments will be carried out in the range of 1 to 10 atm and the gap size of 0.5, 1, 2 mm. The CHF will be detected using 66 type

  15. Ex-Vessel Core Melt Modeling Comparison between MELTSPREAD-CORQUENCH and MELCOR 2.1

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Farmer, Mitchell [Argonne National Lab. (ANL), Argonne, IL (United States); Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-03-01

    System-level code analyses by both United States and international researchers predict major core melting, bottom head failure, and corium-concrete interaction for Fukushima Daiichi Unit 1 (1F1). Although system codes such as MELCOR and MAAP are capable of capturing a wide range of accident phenomena, they currently do not contain detailed models for evaluating some ex-vessel core melt behavior. However, specialized codes containing more detailed modeling are available for melt spreading such as MELTSPREAD as well as long-term molten corium-concrete interaction (MCCI) and debris coolability such as CORQUENCH. In a preceding study, Enhanced Ex-Vessel Analysis for Fukushima Daiichi Unit 1: Melt Spreading and Core-Concrete Interaction Analyses with MELTSPREAD and CORQUENCH, the MELTSPREAD-CORQUENCH codes predicted the 1F1 core melt readily cooled in contrast to predictions by MELCOR. The user community has taken notice and is in the process of updating their systems codes; specifically MAAP and MELCOR, to improve and reduce conservatism in their ex-vessel core melt models. This report investigates why the MELCOR v2.1 code, compared to the MELTSPREAD and CORQUENCH 3.03 codes, yield differing predictions of ex-vessel melt progression. To accomplish this, the differences in the treatment of the ex-vessel melt with respect to melt spreading and long-term coolability are examined. The differences in modeling approaches are summarized, and a comparison of example code predictions is provided.

  16. One dimensional CdS nanowire@TiO2 nanoparticles core-shell as high performance photocatalyst for fast degradation of dye pollutants under visible and sunlight irradiation.

    Science.gov (United States)

    Arabzadeh, Abbas; Salimi, Abdollah

    2016-10-01

    In this study, one-dimensional CdS nanowires@TiO2 nanoparticles core-shell structures (1D CdS NWs@TiO2 NPs) were synthesized by a facile wet chemical-solvothermal method. The different aspects of the properties of CdS NWs@TiO2 NPs were surveyed by using a comprehensive range of characterization techniques including X-ray diffraction (XRD), Fourier transform infrared spectroscopy (FTIR), UV-vis spectroscopy, scanning electron microscopy (SEM), fluorescence spectroscopy, energy dispersive X-ray spectroscopy (EDX), Cyclic Voltammetry (CV) and amperometry. The as-prepared nanostructure was applied as an effective photocatalyst for degradation of methyl orange (MO), methylene blue (MB) and rhodamine B (Rh B) under visible and sunlight irradiation. The results indicated significantly enhanced photocatalytic activity of CdS NWs@TiO2 NPs for degradation of MO, MB and Rh B compared to CdS NWs. The enhanced photocatalytic activity could be attributed to the enhanced sunlight absorbance and the efficient charge separation of the formed heterostructure between CdS NWs and TiO2. The results showed that MO, Rh B and MB were almost completely degraded after 2, 2 and 3min of exposure to sunlight, respectively; while under visible light irradiation (3W blue LED lamp) the dyes were decomposed with less half degradation rate. The catalytic activity was retained even after three degradation cycles of organic dyes, demonstrating that the proposed nanocomposite can be effectively used as efficient photocatalyst for removal of environmental pollutions caused by organic dyes under sunlight irradiation and it could be an important addition to the field of wastewater treatment. We hope the present study may open a new window of such 1-D semiconductor nanocomposites to be used as visible light photocatalysts in the promising field of organic dyes degradation. Copyright © 2016 Elsevier Inc. All rights reserved.

  17. OECD MCCI project 2-D Core Concrete Interaction (CCI) tests : CCI-3 test data report-thermalhydraulic results. Rev. 0 October 15, 2005.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of a third long-term 2-D Core-Concrete Interaction (CCI) experiment designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-3 experiment, which was conducted on September 22, 2005. Test specifications for CCI-3 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 375

  18. OECD MCCI 2-D Core Concrete Interaction (CCI) tests : CCI-2 test data report-thermalhydraulic results, Rev. 0 October 15, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-2 experiment, which was conducted on August 24, 2004. Test specifications for CCI-2 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  19. Prevention and investigations of core degradation in case of beyond design accidents of the 2400 MWTH gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Bertrand, F.; Gatin, V.; Bentivoglio, F.; Gueneau, C.

    2011-01-01

    The present paper deals with studies carried out to assess the ability of the core of the Gas Fast Reactor (GFR) to withstand beyond design accidents. The work presented here is aimed at simulating the behaviour of this core by using analytical models whose input parameters are calculated with the CATHARE2 code. Among possible severe accident initiators, the Unprotected Loss Of Coolant Accident (ULOCA of 3 Inches diameter) is investigated in detail in the paper with CATHARE2. Additionally, a simplified pessimistic assessment of the effect of a postulated power excursion that could result from the failure of prevention provisions is presented. (author)

  20. Safety Strategy of JSFR establishing In-Vessel Retention of Core Disruptive Accident

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu

    2013-01-01

    Coolability of debris bed was confirmed by debris bed temperature analysis coupled with the cooling system, according to the following material relocation scenario. → Case 1: Upward ejection in Transition Phase to cause shutdown. → Case 2: Early downward ejection of fuel through CRGT. → Case 3: Whole fuel accumulates on the core catcher (bounding). The flow reversal of a primary coolant loop of the two loop system of the JSFR which is caused by possible imbalance between two DHRS loops increase the flow in RV. Helpful for long-term cooling

  1. Analysis of coolability of the control rods of a Savannah River Site production reactor with loss of normal forced convection cooling

    International Nuclear Information System (INIS)

    Easterling, T.C.; Hightower, N.T.; Smith, D.C.; Amos, C.N.

    1992-01-01

    An analytical study of the coolability of the control rods in the Savannah River Site (SRS) K-Production Reactor under conditions of loss of normal forced convection cooling has been performed. The study was performed as part of the overall safety analysis of the reactor supporting its restart. The analysis addresses the buoyancy-driven flow over the control rods that occurs when forced cooling is lost, and the limit of critical heat flux that sets the acceptance criteria for the study. The objective of the study is to demonstrate that the control rods will remain cooled at powers representative of those anticipated for restart of the reactor. The study accomplishes this objective with a very tractable simplified analysis for the modest restart power. In addition, a best-estimate calculation is performed, and the results are compared to results from sub-scale scoping experiments. 5 refs

  2. Fabrication of the novel core-shell MCM-41@mTiO{sub 2} composite microspheres with large specific surface area for enhanced photocatalytic degradation of dinitro butyl phenol (DNBP)

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Xiao-Na; Wang, Hui-Long, E-mail: hlwang@dlut.edu.cn; Li, Zhen-Duo; Huang, Zhi-Qiang; Qi, Hui-Ping; Jiang, Wen-Feng

    2016-05-30

    Graphical abstract: The mesoporous MCM-41@mTiO{sub 2} composite microspheres with core/shell structure, well-crystallized mesoporous TiO{sub 2} layer, high specific surface, large pore volume and excellent photocatalytic activity were synthesized by combining sol-gel and simple hydrothermal treatment. - Highlights: • The mesoporous MCM-41@mTiO{sub 2} composite was synthesized successfully. • The composite was facilely prepared by combining sol-gel and hydrothermal method. • The composite exhibited high photocatalytic degradation activity for DNBP. • The composite photocatalyst has excellent reproducibility. - Abstract: The mesoporous MCM-41@mTiO{sub 2} core-shell composite microspheres were synthesized successfully by combining sol-gel and simple hydrothermal treatment. The morphology and microstructure characteristics of the synthesized materials were characterized by scanning electron microscopy (SEM), transmission electron microscopy (TEM), N{sub 2} adsorption-desorption measurements, X-ray powder diffraction (XRD), UV–vis diffuse reflectance spectra (UV–vis/DRS) and Fourier transform infrared spectroscopy (FT-IR). The results indicate that the composite material possesses obvious core/shell structure, a pure mesoporous and well-crystallized TiO{sub 2} layer (mTiO{sub 2}), high specific surface area (316.8 m{sup 2}/g), large pore volume (0.42 cm{sup 3}/g) and two different pore sizes (2.6 nm and 11.0 nm). The photocatalytic activity of the novel MCM-41@mTiO{sub 2} composite was evaluated by degrading 2-sec-butyl-4,6-dinitrophenol (DNBP) in aqueous suspension under UV and visible light irradiation. The results were compared with commercial anatase TiO{sub 2} and Degussa P25 and the enhanced degradation were obtained with the synthesized MCM-41@mTiO{sub 2} composite under the same conditions, which meant that this material can serve as an efficient photocatalyst for the degradation of hazardous organic pollutants in wastewaters.

  3. OECD MMCI 2-D Core Concrete Interaction (CCI) tests : CCCI-1 test data report-thermalhydraulic results. Rev 0 January 31, 2004.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division)

    2011-05-23

    The Melt Attack and Coolability Experiments (MACE) program addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten coreconcrete interactions under both wet and dry cavity conditions. Achievement of these two program objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of satisfying these objectives, the Management Board (MB) approved the conduct of two long-term 2-D Core-Concrete Interaction (CCI) experiments designed to provide information in several areas, including: (i) lateral vs. axial power split during dry core-concrete interaction, (ii) integral debris coolability data following late phase flooding, and (iii) data regarding the nature and extent of the cooling transient following breach of the crust formed at the melt-water interface. This data report provides thermal hydraulic test results from the CCI-1 experiment, which was conducted on December 19, 2003. Test specifications for CCI-1 are provided in Table 1-1. This experiment investigated the interaction of a fully oxidized 400 kg

  4. Importance of the in and ex-vessel corium coolability in case of severe accident for the French PWRs. Some views from L2 PSA and perspectives

    International Nuclear Information System (INIS)

    Raimond, E.; Caroli, C.; Meignen, R.; Rahni, N.; Laurent, B.

    2011-01-01

    In the case of a severe accident on a NPP leading to core degradation after a default in the core cooling as during the accident of Three Mile Island (TMI2), the most efficient way to stop the accident progression would be the in-vessel water injection if a specific mean is available. The TMI2 accident has shown that the accident can be stopped and that the corium, even if highly degraded, can be cooled, but no one can generalize the TMI2 accident termination to all situations. The present paper aims at presenting the situation for the French operated PWRs and is mainly based on the IRSN experience in level 2 probabilistic safety assessment (L2 PSA) development for this type of reactor. It tries to highlight the benefit that could be obtained from a better understanding of the corium cooling phenomenology, including both possible positive and negative effects. Three main negative effects of in-vessel flooding have to be taken into account in a L2 PSA for a PWR: an increase of the hydrogen production rate, a risk of in-vessel pressure increase and the development of conditions for steam explosion. L2 PSAs in France have now reached a certain maturity allowing raising some more precise issues, but for the issues presented in this paper, some progress from the research-development and the simulation tools (mainly the ASTEC integral code) are still necessary to support decision-making

  5. Proceedings of the Second OECD (NEA) CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions

    International Nuclear Information System (INIS)

    1992-01-01

    The Second CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions was held at Kernforschungszentrum Karlsruhe, Germany on April 1-3, 1992. The status and progress in this field of severe reactor accidents were discussed from researchers around the world including participants from Russia and the Czech and Slovak Federal Republic. The contributions concentrated on two main topics. The first topic is the 'classical' core debris-concrete interaction, both experimental and theoretical. Integral effects and separate effects were addressed in thermal hydraulics and heat transfer, material interaction, and aerosol release during concrete erosion, with some applications to prototypical nuclear power plants. The second topic gaining more and more interest is the possibility of controlling and ending the erosion of the concrete by spreading of the core melt, and/or achieving coolability by the addition of water. In the final session it was concluded that considerable progress has been made in understanding and modelling the important phenomena. For the first topic a broad and generally sufficient experimental data base is existing, allowing further improvement qualification of the theoretical models which at present give reasonable agreement with the most important experimental data. A validation matrix is recommended for final validation of the codes. With respect to fission product release during MCCI measurements show that the releases are significantly less than previously estimated. The relatively new topic of melt coolability deserves further investigations which are already underway at different places or international coordinated efforts

  6. Proceedings of the Second OECD (NEA) CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-07-01

    The Second CSNI Specialist Meeting on Molten Core Debris-Concrete Interactions was held at Kernforschungszentrum Karlsruhe, Germany on April 1-3, 1992. The status and progress in this field of severe reactor accidents were discussed from researchers around the world including participants from Russia and the Czech and Slovak Federal Republic. The contributions concentrated on two main topics. The first topic is the 'classical' core debris-concrete interaction, both experimental and theoretical. Integral effects and separate effects were addressed in thermal hydraulics and heat transfer, material interaction, and aerosol release during concrete erosion, with some applications to prototypical nuclear power plants. The second topic gaining more and more interest is the possibility of controlling and ending the erosion of the concrete by spreading of the core melt, and/or achieving coolability by the addition of water. In the final session it was concluded that considerable progress has been made in understanding and modelling the important phenomena. For the first topic a broad and generally sufficient experimental data base is existing, allowing further improvement qualification of the theoretical models which at present give reasonable agreement with the most important experimental data. A validation matrix is recommended for final validation of the codes. With respect to fission product release during MCCI measurements show that the releases are significantly less than previously estimated. The relatively new topic of melt coolability deserves further investigations which are already underway at different places or international coordinated efforts.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  8. OECD MCCI project long-term 2-D molten core concrete interaction test design report, Rev. 0. September 30, 2002

    International Nuclear Information System (INIS)

    Farmer, M.T.; Kilsdonk, D.J.; Lomperski, S.; Aeschliman, R.W.; Basu, S.

    2011-01-01

    The Melt Attack and Coolability Experiments (MACE) program at Argonne National Laboratory addressed the issue of the ability of water to cool and thermally stabilize a molten core-concrete interaction when the reactants are flooded from above. These tests provided data regarding the nature of corium interactions with concrete, the heat transfer rates from the melt to the overlying water pool, and the role of noncondensable gases in the mixing processes that contribute to melt quenching. As a follow-on program to MACE, The Melt Coolability and Concrete Interaction Experiments (MCCI) project is conducting reactor material experiments and associated analysis to achieve the following two technical objectives: (1) resolve the ex-vessel debris coolability issue through a program that focuses on providing both confirmatory evidence and test data for the coolability mechanisms identified in MACE integral effects tests, and (2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs for future plants. In terms of the first program objective, the Small-Scale Water Ingression and Crust Strength (SSWICS) test series has been initiated to provide fundamental information on the ability of water to ingress into cracks and fissures that form in the debris during quench, thereby augmenting the otherwise conduction-limited heat transfer process. A test plan for Melt Eruption Separate Effects Tests (MESET) has also been developed to provide information on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions. In terms of the second program objective, the project Management Board (MB) has approved startup activities required to carry out

  9. Severe accident research in the core degradation area: An example of effective international cooperation between the European Union (EU) and the Commonwealth of Independent States (CIS) by the International Science and Technology Center

    Energy Technology Data Exchange (ETDEWEB)

    Bottomley, D., E-mail: paul.bottomley@ec.europa.eu [ITU Institut fuer Transurane, PO box 2340, 76125 Karlsruhe (Germany); Stuckert, J.; Hofmann, P. [KIT Campus Nord, Hermann-von-Helmholtz Pl. 1, 76344 Eggenstein-Leopoldshafen (Germany); Tocheny, L. [ISTC Krasnoproletarskaya 32-34, PO Box 20, 127473 Moscow (Russian Federation); Hugon, M. [European Commission DG - Research and Tech. Development, Sq. de Meeus, B-1049 Brussels (Belgium); Journeau, C. [CEA, DEN, Cadarache, F13108 St Paul lez Durance (France); Clement, B. [IRSN PSN-RES/SAG Cadarache, BP3 F13115, St Paul lez Durance (France); Weber, S. [GRS Muenchen, Thermal Hydraulics Div., Garching 85748,Germany (Germany); Guentay, S. [PSI NES/LTH OHSA C11, 5232 Villigen (Switzerland); Hozer, Z. [AEKI Fuel Department, P.O. Box 49, Budapest H-1525 (Hungary); Herranz, L. [CIEMAT, Energy -Nuclear Fission Division, Complutense 40, 28040 Madrid (Spain); Schumm, A. [EDF - R and D, SINETICS, Avenue du General de Gaulle 1, Clamart 92140 (France); Oriolo, F. [Pisa University, Ing. Mecc. Nucl. Prod., Largo Lazarino 2, Pisa 56126 (Italy); Altstadt, E. [HZDR Structural Matls, Rossendorf, Postfach 51 01 19, 01314 Dresden (Germany); Krause, M. [AECL - Reactor Safety, Chalk River, Ontario, Canada K0J 1J0 (Canada); Fischer, M. [AREVA NP GMBH, Dept. PEPA-G, 91058 Erlangen (Germany); Khabensky, V.B. [Alexandrov Institute of Technologies (NITI), Sosnovy Bor (Russian Federation); Bechta, S.V. [Kungliga Tekniska Hoegskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91 Stockholm (Sweden); Veshchunov, M.S. [Nuclear Safety Institute (IBRAE), Russian Academy of Sciences, 52 B. Tulskaya, Moscow 115191 (Russian Federation); Palagin, A.V. [KIT Campus Nord, Hermann-von-Helmholtz Pl. 1, 76344 Eggenstein-Leopoldshafen (Germany); and others

    2012-11-15

    temperature; (2) Reactor Core Degradation; a modelling project simulating the fuel rod degradation and loss of geometry from IBRAE, Moscow; (3) METCOR projects from NITI, St. Petersburg on the interaction of core melt with reactor vessel steel; (4) INVECOR project, NNE Kurchatov City, Kazakhstan; this is a large-scale facility to examine the vessel steel retention of 60 kg corium during the decay heat; and finally, (5) CORPHAD and PRECOS projects, NITI, St. Petersburg undertook a systematic examination of refractory ceramics relevant to in-vessel and ex-vessel coria, particularly examining poorly characterised, limited data or experimentally difficult systems.

  10. Degradation and leaching behaviour of 14C-glufosinate in a silty sand soil. Experiments in outdoor lysimeters with undisturbed soil cores

    International Nuclear Information System (INIS)

    Kubiak, R.

    1996-12-01

    Degradation and leaching behaviour of 14 C-labelled glufosinate in a silty sand soil was investigated in two outdoor lysimeters after repeated application of 12.5 litres/hectare (1/ha) Basta (divided in 7.5 and 5 l/ha respectively). The 14 C-loss during application was 4.8-8.2%. The 14 C-content in the plants (vines and weeds) was 0.3% of that applied at the most. After 130 days, 25.9 and 25.5% of the applied material was found in the soil up to a depth of 40 cm. One year after the first application, this amount was still 18.5 and 18.6%. As a consequence of the renewed spraying, the detected amounts of 14 C were 44.3 and 43.1% some 107 days after the first application in the second experimental year. The additional investigation in lysimeter 2 after 373 days showed a decrease to 33.9%. Most of the detected radioactivity remained in the 0-10 cm soil layer. At the end of the experiment, the amount of 14 C in the 30-40 cm layer was 0.5%. The total residues in the 0-10 cm soil layer were less than 1 mg/kg at all dates of sampling, and only a small amount still represented the free acid of the active ingredient. The average values were 0.05 mg/kg after 130 days, 0.01 mg/kg after 363 days and 0.09 mg/kg at the following date of sampling. In the spring of the following year, no residues of the free acid were detectable. The radioactivity in the percolate amounted to a maximum of 0.11% of that applied and in no case represented the free acid of the ammonium salt. (author)

  11. Degradation and leaching behaviour of {sup 14}C-glufosinate in a silty sand soil. Experiments in outdoor lysimeters with undisturbed soil cores

    Energy Technology Data Exchange (ETDEWEB)

    Kubiak, R

    1996-12-01

    Degradation and leaching behaviour of {sup 14}C-labelled glufosinate in a silty sand soil was investigated in two outdoor lysimeters after repeated application of 12.5 litres/hectare (1/ha) Basta (divided in 7.5 and 5 l/ha respectively). The {sup 14}C-loss during application was 4.8-8.2%. The {sup 14}C-content in the plants (vines and weeds) was 0.3% of that applied at the most. After 130 days, 25.9 and 25.5% of the applied material was found in the soil up to a depth of 40 cm. One year after the first application, this amount was still 18.5 and 18.6%. As a consequence of the renewed spraying, the detected amounts of {sup 14}C were 44.3 and 43.1% some 107 days after the first application in the second experimental year. The additional investigation in lysimeter 2 after 373 days showed a decrease to 33.9%. Most of the detected radioactivity remained in the 0-10 cm soil layer. At the end of the experiment, the amount of {sup 14}C in the 30-40 cm layer was 0.5%. The total residues in the 0-10 cm soil layer were less than 1 mg/kg at all dates of sampling, and only a small amount still represented the free acid of the active ingredient. The average values were 0.05 mg/kg after 130 days, 0.01 mg/kg after 363 days and 0.09 mg/kg at the following date of sampling. In the spring of the following year, no residues of the free acid were detectable. The radioactivity in the percolate amounted to a maximum of 0.11% of that applied and in no case represented the free acid of the ammonium salt. (author)

  12. Core failure accident pathways and ways to control it

    International Nuclear Information System (INIS)

    Mayinger, F.

    1982-01-01

    In the German Risk Study accidents are assumed to result in core meltdown whenever the criteria spelt out in the guidelines of the Advisory Committee on Reactor Safeguards are no longer met. This assumption must be seen in the light of an earlier state of the art in which no detailed information could be obtained about intermediate stages in emergency core cooling systems working according to permit up to the complete failure of all heat removal systems. However, experimental studies and theoretical analyses conducted over the past few years have advanced the state of the art such that it is now possible to predict with considerably more physical reality the behavior of a core in a loss-of-coolant accident. These findings are not only based on calculations, but also on the results of experiments in large facilities allowing direct comparisons to be made with conditions in nuclear power plants. Studies of the effects of systems failures both in major leakages and in the small leakages regarded to be much more dangerous show much more favorable conditions with respect to core coolability than had to be anticipated on the basis of earlier assumptions. This also implies that it would neither be necessary nor meaningful to reinforce emergency core cooling systems. Instead, it is much more important, besides having technically highly qualified and thoroughly trained operating crews, to inform those crews reliably of the hydrodynamic and thermodynamic state of the primary system, especially the core. (orig.) [de

  13. GFR fuel and core pre-conceptual design studies

    International Nuclear Information System (INIS)

    Chauvin, N.; Ravenet, A.; Lorenzo, D.; Pelletier, M.; Escleine, J.M.; Munoz, I.; Bonnerot, J.M.; Malo, J.Y.; Garnier, J.C.; Bertrand, F.; Bosq, J.C.

    2007-01-01

    The revision of the GFR core design - plate type - has been undertaken since previous core presented at Global'05. The self-breeding searched for has been achieved with an optimized design ('12/06 E'). The higher core pressure drop was a matter of concern. First of all, the core coolability in natural circulation for pressurized conditions has been studied and preliminary plant transient calculations have been performed. The design and safety criteria are met but no more margin remains. The project is also addressing the feasibility and the design of the fuel S/A. The hexagonal shape together with the principle of closed S/A (wrapper tube) is kept. Ceramic plate type fuel element combines a high enough core power density (minimization of the Pu inventory) and plutonium and minor actinides recycling capabilities. Innovative for many aspects, the fuel element is central to the GFR feasibility. It is supported already by a significant R and D effort also applicable to a pin concept that is considered as the other fuel element of interest. This combination of fuel/core feasibility and performance analysis, safety dispositions and performances analysis will compose the 'GFR preliminary feasibility' which is a project milestone at the end of the year 2007. (authors)

  14. OECD/MCCI 2-D Core Concrete Interaction (CCI) tests : final report February 28, 2006.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Lomperski, S.; Kilsdonk, D. J.; Aeschlimann, R. W.; Basu, S. (Nuclear Engineering Division); (NRC)

    2011-05-23

    Although extensive research has been conducted over the last several years in the areas of Core-Concrete Interaction (CCI) and debris coolability, two important issues warrant further investigation. The first issue concerns the effectiveness of water in terminating a CCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. This safety issue was investigated in the EPRI-sponsored Melt Attack and Coolability Experiments (MACE) program. The approach was to conduct large scale, integral-type reactor materials experiments with core melt masses ranging up to two metric tons. These experiments provided unique, and for the most part repeatable, indications of heat transfer mechanism(s) that could provide long term debris cooling. However, the results did not demonstrate definitively that a melt would always be completely quenched. This was due to the fact that the crust anchored to the test section sidewalls in every test, which led to melt/crust separation, even at the largest test section lateral span of 1.20 m. This decoupling is not expected for a typical reactor cavity, which has a span of 5-6 m. Even though the crust may mechanically bond to the reactor cavity walls, the weight of the coolant and the crust itself is expected to periodically fracture the crust and restore contact with the melt. Although crust fracturing does not ensure that coolability will be achieved, it nonetheless provides a pathway for water to recontact the underlying melt, thereby allowing other debris cooling mechanisms to proceed. A related task of the current program, which is not addressed in this particular report, is to measure crust strength to check the hypothesis that a corium crust would not be strong enough to sustain melt/crust separation in a plant accident. The second important issue concerns long-term, two-dimensional concrete ablation by a prototypic core oxide melt. As discussed by Foit the existing

  15. An assessment of Class-9 (core-melt) accidents for PWR dry-containment systems

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Saito, M.

    1981-01-01

    The phenomenology of core-melt accidents in dry containments was examined for the purpose of identifying the margins of safety in such Class-9 situations. The scale (geometry) effects appear to crucially limit the extent (severity) of steam explosions. This together with the established reduced explosivity of the corium-A/water system, and the inherently high capability of dry containments (redinforced concrete, and shields in some cases, seismic design etc.) lead to the conclusion that failure due to steam explosions may be considered essentially incredible. These premixture scaling considerations also impact ultimate debris disposition and coolability and need additional development. A water-flooded reactor cavity would have beneficial effects in limiting (but not necessarily eliminating) melt-concrete interactions. Independently of the initial degree of quenching and/or scale of fragmentation, mechanisms exist that drive the system towards ultimate stability (coolability). Additional studies, with intermediate-scale prototypic materials are recommended to better explore these mechanisms. Containment heat removal systems must provide the crucial capability of mitigating such accidents. Passive systems should be explored and assessed against currently available and/or improved active systems taking into account the rather loose time constraints required for activation. It appears that containment margins for accommodating the hydrogen problem are limited. This problem appears to stand out not only in terms of potential consequences but also in terms of lack of any readily available and clear cut solutions at this time. (orig.)

  16. Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core Concrete Interaction

    International Nuclear Information System (INIS)

    Robb, Kevin R; Farmer, Mitchell; Francis, Matthew W

    2015-01-01

    Lower head failure and corium concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH.

  17. Ex-vessel core catcher design requirements and preliminary concepts evaluation

    International Nuclear Information System (INIS)

    Friedland, A.J.; Tilbrook, R.W.

    1974-01-01

    As part of the overall study of the consequences of a hypothetical failure to scram following loss of pumping power, design requirements and preliminary concepts evaluation of an ex-vessel core catcher (EVCC) were performed. EVCC is the term applied to a class of devices whose primary objective is to provide a stable subcritical and coolable configuration within containment following a postulated accident in which it is assumed that core debris has penetrated the Reactor Vessel and Guard Vessel. Under these assumed conditions a set of functional requirements were developed for an EVCC and several concepts were evaluated. The studies were specifically directed toward the FFTF design considering the restraints imposed by the physical design and construction of the FFTF plant

  18. Simulant - water experiments to characterize the debris bed formed in severe core melt accidents

    International Nuclear Information System (INIS)

    Mathai, Amala M.; Anandan, J.; Sharma, Anil Kumar; Murthy, S.S.; Malarvizhi, B.; Lydia, G.; Das, Sanjay Kumar; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Molten Fuel Coolant Interaction (WO) and debris bed configuration on the core catcher plate assumes importance in assessing the Post Accident Heat Removal (PARR) of a heat generating debris bed. The key factors affecting the coolability of the debris bed are the bed porosity, morphology of the fragmented particles, degree of spreading/heaping of the debris on the core catcher and the fraction of lump formed. Experiments are conducted to understand the fragmentation kinetics and subsequent debris bed formation of molten woods metal in water at interface temperatures near the spontaneous nucleation temperature of water. Morphology of the debris particles is investigated to understand the fragmentation mechanisms involved. The spreading behavior of the debris on the catcher plate and the particle size distribution are presented for 5 kg and 10 kg melt inventories. Porosity of the undisturbed bed on the catcher plate is evaluated using a LASER sensor technique. (author)

  19. TMI-2 core examination plan

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. Proteomics Core

    Data.gov (United States)

    Federal Laboratory Consortium — Proteomics Core is the central resource for mass spectrometry based proteomics within the NHLBI. The Core staff help collaborators design proteomics experiments in a...

  1. Proposal for computer investigation of LMFBR core meltdown accidents

    International Nuclear Information System (INIS)

    Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

    1974-01-01

    The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

  2. Experimental Study for Effects of the Stud shape of the Core Catcher System

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Seo, Gwang Hyeok; Shin, Doyoung; Jeun, Gyoodong; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2015-05-15

    In preparation of potential severe accidents, a nuclear power plant is equipped with diverse systems of engineering safety features or mitigation system dedicated to the severe accidents conditions. As a common strategy, a number of nuclear power plants adopt the in-vessel retention (IVR) and/or external reactor vessel cooling (ERVC) strategies. With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The core catcher system is for Ex-vessel in the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while keeping coolable geometry in case that the RPV failure occurs. The system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. As shown in Fig 1, the engineered cooling channel of the ex-vessel core catcher was adopted to remove sensible heat and decay heat of the molten corium using cooling water flooded from the In-Containment Refueling Water Storage Tank (IRWST) by gravity. A large number of studs are placed in the cooling channel to support the core catcher body. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the temperature distribution and overall coolability of the system. Thus, it is of importance to investigate the effects of studs on the coolability of the core catcher system. In the current research, to evaluate the effect of a stud on the streamline and natural convective boiling performance, numerical and experimental approaches were taken. As a part of numerical approach, CFD simulation using ANSYS/FLUENT was carried out. The objective was to predict disturbance of the streamline and temperature distribution due to the interference of the studs. Through the CFD

  3. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2008-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  4. Transformer core

    NARCIS (Netherlands)

    Mehendale, A.; Hagedoorn, Wouter; Lötters, Joost Conrad

    2010-01-01

    A transformer core includes a stack of a plurality of planar core plates of a magnetically permeable material, which plates each consist of a first and a second sub-part that together enclose at least one opening. The sub-parts can be fitted together via contact faces that are located on either side

  5. Flow Boiling on a Downward-Facing Inclined Plane Wall of Core Catcher

    International Nuclear Information System (INIS)

    Kim, Hyoung Tak; Bang, Kwang Hyun; Suh, Jung Soo

    2013-01-01

    In order to investigate boiling behavior on downward-facing inclined heated wall prior to the CHF condition, an experiment was carried out with 1.2 m long rectangular channel, inclined by 10 .deg. from the horizontal plane. High speed video images showed that the bubbles were sliding along the heated wall, continuing to grow and combining with the bubbles growing at their nucleation sites in the downstream. These large bubbles continued to slide along the heated wall and formed elongated slug bubbles. Under this slug bubble thin liquid film layer on the heated wall was observed and this liquid film prevents the wall from dryout. The length, velocity and frequency of slug bubbles sliding on the heated wall were measured as a function of wall heat flux and these parameters were used to develop wall boiling model for inclined, downward-facing heated wall. One approach to achieve coolable state of molten core in a PWR-like reactor cavity during a severe accident is to retain the core melt on a so-called core catcher residing on the reactor cavity floor after its relocation from the reactor pressure vessel. The core melt retained in the core catcher is cooled by water coolant flowing in an inclined cooling channel underneath as well as the water pool overlaid on the melt layer. Two-phase flow boiling with downward-facing heated wall such as this core catcher cooling channel has drawn a special attention because this orientation of heated wall may reach boiling crisis at lower heat flux than that of a vertical or upward-facing heated wall. Nishikawa and Fujita, Howard and Mudawar, Qiu and Dhir have conducted experiments to study the effect of heater orientation on boiling heat transfer and CHF. SULTAN experiment was conducted to study inclined large-scale structure coolability by water in boiling natural convection. In this paper, high-speed visualization of boiling behavior on downward-facing heated wall inclined by 10 .deg. is presented and wall boiling model for the

  6. Core lifter

    Energy Technology Data Exchange (ETDEWEB)

    Pavlov, N G; Edel' man, Ya A

    1981-02-15

    A core lifter is suggested which contains a housing, core-clamping elements installed in the housing depressions in the form of semirings with projections on the outer surface restricting the rotation of the semirings in the housing depressions. In order to improve the strength and reliability of the core lifter, the semirings have a variable transverse section formed from the outside by the surface of the rotation body of the inner arc of the semiring aroung the rotation axis and from the inner a cylindrical surface which is concentric to the outer arc of the semiring. The core-clamping elements made in this manner have the possibility of freely rotating in the housing depressions under their own weight and from contact with the core sample. These semirings do not have weakened sections, have sufficient strength, are inserted into the limited ring section of the housing of the core lifter without reduction in its through opening and this improve the reliability of the core lifter in operation.

  7. Organic matter degradation in Chilean sediments - following nature's own degradation experiment

    DEFF Research Database (Denmark)

    Langerhuus, Alice Thoft; Niggemann, Jutta; Lomstein, Bente Aagaard

    ORGANIC MATTER DEGRADATION IN CHILEAN SEDIMENTS – FOLLOWING NATURE’S OWN DEGRADATION EXPERIMENT Degradation of sedimentary organic matter was studied at two stations from the shelf of the Chilean upwelling region. Sediment cores were taken at 1200 m and 800 m water depth and were 4.5 m and 7.5 m...... in length, respectively. The objective of this study was to assess the degradability of the organic matter from the sediment surface to the deep sediments. This was done by analysing amino acids (both L- and D-isomers) and amino sugars in the sediment cores, covering a timescale of 15.000 years. Diagenetic...... indicators (percentage of carbon and nitrogen present as amino acid carbon and nitrogen, the ratio between a protein precursor and its non-protein degradation product and the percentage of D-amino acids) revealed ongoing degradation in these sediments, indicating that microorganisms were still active in 15...

  8. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  9. Ice Cores

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Records of past temperature, precipitation, atmospheric trace gases, and other aspects of climate and environment derived from ice cores drilled on glaciers and ice...

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

    International Nuclear Information System (INIS)

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

    1980-08-01

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

  11. Birefringent hollow core fibers

    DEFF Research Database (Denmark)

    Roberts, John

    2007-01-01

    Hollow core photonic crystal fiber (HC-PCF), fabricated according to a nominally non-birefringent design, shows a degree of un-controlled birefringence or polarization mode dispersion far in excess of conventional non polarization maintaining fibers. This can degrade the output pulse in many...... applications, and places emphasis on the development of polarization maintaining (PM) HC-PCF. The polarization cross-coupling characteristics of PM HC-PCF are very different from those of conventional PM fibers. The former fibers have the advantage of suffering far less from stress-field fluctuations...... and an increased overlap between the polarization modes at the glass interfaces. The interplay between these effects leads to a wavelength for optimum polarization maintenance, lambda(PM), which is detuned from the wavelength of highest birefringence. By a suitable fiber design involving antiresonance of the core...

  12. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  13. Core microbiomes for sustainable agroecosystems.

    Science.gov (United States)

    Toju, Hirokazu; Peay, Kabir G; Yamamichi, Masato; Narisawa, Kazuhiko; Hiruma, Kei; Naito, Ken; Fukuda, Shinji; Ushio, Masayuki; Nakaoka, Shinji; Onoda, Yusuke; Yoshida, Kentaro; Schlaeppi, Klaus; Bai, Yang; Sugiura, Ryo; Ichihashi, Yasunori; Minamisawa, Kiwamu; Kiers, E Toby

    2018-05-01

    In an era of ecosystem degradation and climate change, maximizing microbial functions in agroecosystems has become a prerequisite for the future of global agriculture. However, managing species-rich communities of plant-associated microbiomes remains a major challenge. Here, we propose interdisciplinary research strategies to optimize microbiome functions in agroecosystems. Informatics now allows us to identify members and characteristics of 'core microbiomes', which may be deployed to organize otherwise uncontrollable dynamics of resident microbiomes. Integration of microfluidics, robotics and machine learning provides novel ways to capitalize on core microbiomes for increasing resource-efficiency and stress-resistance of agroecosystems.

  14. Safety analysis of the topaz behavior during irradiation, its effect on the core performance and the in-core fuel management strategy

    International Nuclear Information System (INIS)

    Khalil, M.Y.; Belal, M.G.

    2006-01-01

    The topaz is a natural gem stones which collect color centers when irradiated with fast neutrons and transformed into a colorful stones called topaz. The objective of this paper is to detail the safety analysis performed to assure the safety measures of the topaz mass production and farther shows an indirect estimated measurement of the safety related parameters. Analysis has been performed for all the irradiation positions nominated for topaz production and this paper present experimental verification performed for the position of the highest influence where all other positions have lower influences and showed the same safety features and agreement between calculations and measurements. On the other hand it was necessary to show that no hot spots and no cooling problems would rise as a result of irradiation. The heat energy dissipation in the topaz boxes is important from the reactor core coolability side as well as from the view point of the quality of the product. Moreover the paper describes the administrative procedure to limit the reactivity insertion rate of any box to less than 10 pcm/sec. The effect of the topaz boxes presence on the accumulated fuel burn up has been calculated, and recommendations concerning the in-core fuel management strategy has been reviewed. (authors)

  15. Development of severe accident analysis code - A study on the molten core-concrete interaction under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Chang Hyun; Lee, Byung Chul; Huh, Chang Wook; Kim, Doh Young; Kim, Ju Yeul [Seoul National University, Seoul (Korea, Republic of)

    1996-07-01

    The purpose of this study is to understand the phenomena of the molten core/concrete interaction during the hypothetical severe accident, and to develop the model for heat transfer and physical phenomena in MCCIs. The contents of this study are analysis of mechanism in MCCIs and assessment of heat transfer models, evaluation of model in CORCON code and verification in CORCON using SWISS and SURC Experiments, and 1000 MWe PWR reactor cavity coolability, and establishment a model for prediction of the crust formation and temperature of melt-pool. The properties and flow condition of melt pool covering with the conditions of severe accident are used to evaluate the heat transfer coefficients in each reviewed model. Also, the scope and limitation of each model for application is assessed. A phenomenological analysis is performed with MELCOR 1.8.2 and MELCOR 1.8.3 And its results is compared with corresponding experimental reports of SWISS and SURC experiments. And the calculation is performed to assess the 1000 MWe PWR reactor cavity coolability. To improve the heat transfer model between melt-pool and overlying coolant and analyze the phase change of melt-pool, 2 dimensional governing equations are established using the enthalpy method and computational program is accomplished in this study. The benchmarking calculation is performed and its results are compared to the experiment which has not considered effects of the coolant boiling and the gas injection. Ultimately, the model shall be developed for considering the gas injection effect and coolant boiling effect. 66 refs., 10 tabs., 29 refs. (author)

  16. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  17. Core BPEL

    DEFF Research Database (Denmark)

    Hallwyl, Tim; Højsgaard, Espen

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

  18. Regulating the 20S Proteasome Ubiquitin-Independent Degradation Pathway

    Directory of Open Access Journals (Sweden)

    Gili Ben-Nissan

    2014-09-01

    Full Text Available For many years, the ubiquitin-26S proteasome degradation pathway was considered the primary route for proteasomal degradation. However, it is now becoming clear that proteins can also be targeted for degradation by the core 20S proteasome itself. Degradation by the 20S proteasome does not require ubiquitin tagging or the presence of the 19S regulatory particle; rather, it relies on the inherent structural disorder of the protein being degraded. Thus, proteins that contain unstructured regions due to oxidation, mutation, or aging, as well as naturally, intrinsically unfolded proteins, are susceptible to 20S degradation. Unlike the extensive knowledge acquired over the years concerning degradation by the 26S proteasome, relatively little is known about the means by which 20S-mediated proteolysis is controlled. Here, we describe our current understanding of the regulatory mechanisms that coordinate 20S proteasome-mediated degradation, and highlight the gaps in knowledge that remain to be bridged.

  19. Enlarged Halden programme group meeting on high burn-up fuel performance, safety and reliability and degradation of in-core materials and water chemistry effects and man-machine systems research. Volume II

    International Nuclear Information System (INIS)

    1999-01-01

    Academy of Sciences, KFKI Atomic Energy Research Institute, the N.V. KEMA, the Netherlands, the Russian Research Centre 'Kurchatov Institute', the Slovakian VUJE - Nuclear Power Plant Research Institute, and from USA: the ABB Combustion Engineering Inc., the Electric Power Research Institute (EPRI), and the General Electric Co. The right to utilise information originating from the research work of the Halden Project is limited to persons and undertakings specifically given this right by one of these Project member organisations. The activities in the area of fuel and materials performance are based on extensive in-reactor measurements. The programmes are expanding in the areas of fuel performance at extended burn-ups, waterside corrosion and material testing in general. Development of in-core instruments is an important activity in support of the experimental programmes. The research programme at the Halden Project addresses the research needs of the nuclear industry in connection with introduction of digital I and C systems in NPPs. The programme provides information supporting design and licensing of upgraded, computer-based control room systems, and demonstrates the benefits of such systems through validation experiments in Halden's experimental research facility, HAMMLAB and pilot installations in NPPs. The Enlarged Halden Programme Group Meeting at Loen, Norway, was arranged to provide an opportunity to present results of work carried out at Halden and within participating organisations, and to encourage comments and impulses related to future Halden Project work. This HPR-351 relates to the fuel and materials part of the meeting and is divided in two volumes, HPR-351 Volume I and HPR-351 Volume II. The corresponding collection of papers in the man-machine area are given in one volume, HPR-352 Volume I. The overall programme of the Loen Enlarged Meeting covering the Fuel and Materials Research is given in the following pages. The papers with denomination HWR have

  20. Core-shell microparticles for protein sequestration and controlled release of a protein-laden core.

    Science.gov (United States)

    Rinker, Torri E; Philbrick, Brandon D; Temenoff, Johnna S

    2017-07-01

    Development of multifunctional biomaterials that sequester, isolate, and redeliver cell-secreted proteins at a specific timepoint may be required to achieve the level of temporal control needed to more fully regulate tissue regeneration and repair. In response, we fabricated core-shell heparin-poly(ethylene-glycol) (PEG) microparticles (MPs) with a degradable PEG-based shell that can temporally control delivery of protein-laden heparin MPs. Core-shell MPs were fabricated via a re-emulsification technique and the number of heparin MPs per PEG-based shell could be tuned by varying the mass of heparin MPs in the precursor PEG phase. When heparin MPs were loaded with bone morphogenetic protein-2 (BMP-2) and then encapsulated into core-shell MPs, degradable core-shell MPs initiated similar C2C12 cell alkaline phosphatase (ALP) activity as the soluble control, while non-degradable core-shell MPs initiated a significantly lower response (85+19% vs. 9.0+4.8% of the soluble control, respectively). Similarly, when degradable core-shell MPs were formed and then loaded with BMP-2, they induced a ∼7-fold higher C2C12 ALP activity than the soluble control. As C2C12 ALP activity was enhanced by BMP-2, these studies indicated that degradable core-shell MPs were able to deliver a bioactive, BMP-2-laden heparin MP core. Overall, these dynamic core-shell MPs have the potential to sequester, isolate, and then redeliver proteins attached to a heparin core to initiate a cell response, which could be of great benefit to tissue regeneration applications requiring tight temporal control over protein presentation. Tissue repair requires temporally controlled presentation of potent proteins. Recently, biomaterial-mediated binding (sequestration) of cell-secreted proteins has emerged as a strategy to harness the regenerative potential of naturally produced proteins, but this strategy currently only allows immediate amplification and re-delivery of these signals. The multifunctional, dynamic

  1. 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

  2. Materials behaviour in PWRs core

    International Nuclear Information System (INIS)

    Barbu, A.; Massoud, J.P.

    2008-01-01

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

  3. Study of heat removal by natural convection from the internal core catcher in PFBR using water model experiments

    International Nuclear Information System (INIS)

    Jasmin Sudha, A.; Punitha, G.; Das, S.K.; Lydia, G.; Murthy, S.S.; Malarvizhi, B.; Harvey, J.; Kannan, S.E.

    2005-01-01

    Full text of publication follows: In the event of a core meltdown accident in a Fast Breeder Reactor, the molten core material settling on the bottom of the main vessel can endanger the structural integrity of the main vessel. In the design of Prototype Fast Breeder Reactor in India, the construction of which is about to commence, a core catcher is provided as the internal core retention device to collect and retain the core debris in a coolable configuration. Heat transfer by natural convection above and below the core catcher plate, in the zone beneath the core support structure is evaluated from water mockup experiments in the 1:4 geometrically scaled setup. These studies were undertaken towards comparison of experimentally measured temperatures at different locations with the numerical results. The core catcher assembly consists of a core catcher plate, a heat shield plate and a chimney. Decay heat from the core debris is simulated by electrical heating of the heat shield plate. An opening is provided in the cover plate to reproduce the situation in the actual accident where the core debris would have breached a part of the core support structure. Experiments were carried out with different heat flux levels prevailing upon the heat shield plate. Temperature monitoring was done at more than 100 locations, distributed both on the solid components and in water. The temperature data was analysed to get the temperature profile at different steady state conditions. Flow visualisation was also carried out using water soluble dye to establish the direction of the convective currents. The captured images show that water flows through the slots provided in the top portion of the chimney in the upward direction as evidenced from the diffusion of dye injected inside the chimney. Both the temperature data and flow visualisation confirm mixing of water through the opening in the core support structure which indicates that natural convection is set up in that zone

  4. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  5. Side core lifter

    Energy Technology Data Exchange (ETDEWEB)

    Edelman, Ya A

    1982-01-01

    A side core lifter is proposed which contains a housing with guide slits and a removable core lifter with side projections on the support section connected to the core receiver. In order to preserve the structure of the rock in the core sample by means of guaranteeing rectilinear movement of the core lifter in the rock, the support and core receiver sections are hinged. The device is equipped with a spring for angular shift in the core-reception part.

  6. Degradation of microbial polyesters.

    Science.gov (United States)

    Tokiwa, Yutaka; Calabia, Buenaventurada P

    2004-08-01

    Microbial polyhydroxyalkanoates (PHAs), one of the largest groups of thermoplastic polyesters are receiving much attention as biodegradable substitutes for non-degradable plastics. Poly(D-3-hydroxybutyrate) (PHB) is the most ubiquitous and most intensively studied PHA. Microorganisms degrading these polyesters are widely distributed in various environments. Although various PHB-degrading microorganisms and PHB depolymerases have been studied and characterized, there are still many groups of microorganisms and enzymes with varying properties awaiting various applications. Distributions of PHB-degrading microorganisms, factors affecting the biodegradability of PHB, and microbial and enzymatic degradation of PHB are discussed in this review. We also propose an application of a new isolated, thermophilic PHB-degrading microorganism, Streptomyces strain MG, for producing pure monomers of PHA and useful chemicals, including D-3-hydroxycarboxylic acids such as D-3-hydroxybutyric acid, by enzymatic degradation of PHB.

  7. Validation of ASTEC core degradation and containment models

    International Nuclear Information System (INIS)

    Kruse, Philipp; Brähler, Thimo; Koch, Marco K.

    2014-01-01

    Ruhr-Universitaet Bochum performed in a German funded project validation of in-vessel and containment models of the integral code ASTEC V2, jointly developed by IRSN (France) and GRS (Germany). In this paper selected results of this validation are presented. In the in-vessel part, the main point of interest was the validation of the code capability concerning cladding oxidation and hydrogen generation. The ASTEC calculations of QUENCH experiments QUENCH-03 and QUENCH-11 show satisfactory results, despite of some necessary adjustments in the input deck. Furthermore, the oxidation models based on the Cathcart–Pawel and Urbanic–Heidrick correlations are not suitable for higher temperatures while the ASTEC model BEST-FIT based on the Prater–Courtright approach at high temperature gives reliable enough results. One part of the containment model validation was the assessment of three hydrogen combustion models of ASTEC against the experiment BMC Ix9. The simulation results of these models differ from each other and therefore the quality of the simulations depends on the characteristic of each model. Accordingly, the CPA FRONT model, corresponding to the simplest necessary input parameters, provides the best agreement to the experimental data

  8. Animal MRI Core

    Data.gov (United States)

    Federal Laboratory Consortium — The Animal Magnetic Resonance Imaging (MRI) Core develops and optimizes MRI methods for cardiovascular imaging of mice and rats. The Core provides imaging expertise,...

  9. The influence of core materials and mix on the performance of a 100 kVA three phase transformer core

    Energy Technology Data Exchange (ETDEWEB)

    Snell, David E-mail: dave.snell@cogent-power.com; Coombs, Alan

    2003-01-01

    Various grades of grain-oriented electrical steel, and the effect of mixing domain refined and non-domain refined materials in the same three phase transformer core have been assessed using a developed computer-based test system. Ball unit domain refined material and non-domain refined material can be successfully mixed in the same core, without degrading performance.

  10. Dryout heat flux and flooding phenomena in debris beds consisting of homogeneous diameter particles

    International Nuclear Information System (INIS)

    Maruyama, Yu; Abe, Yutaka; Yamano, Norihiro; Soda, Kunihisa

    1988-08-01

    Since the TMI-2 accident, which occurred in 1979, necessity of understanding phenomena associated with a severe accident have been recognized and researches have been conducted in many countries. During a severe accident of a light water reactor, a debris bed consisting of the degraded core materials would be formed. Because the debris bed continues to release decay heat, the debris bed would remelt when the coolable geometry is not maintained. Thus the degraded core coolability experiments to investigate the influence of the debris particle diameter and coolant flow conditions on the coolability of the debris bed and the flooding experiments to investigate the dependence of flooding phenomena on the configuration of the debris bed have been conducted in JAERI. From the degraded core coolability experiments, the following conclusions were derived; the coolability of debris beds would be improved by coolant supply into the beds, Lipinski's 1-dimensional model shows good agreement with the measured dryout heat flux for the beds under stagnant and forced flow conditions from the bottom of the beds, and the analytical model used for the case that coolant is fed by natural circulation through the downcomer reproduces the experimental results. And the following conclusions were given from the flooding experiments ; no dependence between bed height and the flooding constant exists for the beds lower than the critical bed height, flooding phenomena of the stratified beds would be dominated by the layer consisting of smaller particles, and the predicted dryout heat flux by the analytical model based on the flooding theory gives underestimation under stagnant condition. (author)

  11. Degradations and Rearrangement Reactions

    Science.gov (United States)

    Zhang, Jianbo

    This section deals with recent reports concerning degradation and rearrangement reactions of free sugars as well as some glycosides. The transformations are classified in chemical and enzymatic ways. In addition, the Maillard reaction will be discussed as an example of degradation and rearrangement transformation and its application in current research in the fields of chemistry and biology.

  12. Core Hunter 3: flexible core subset selection.

    Science.gov (United States)

    De Beukelaer, Herman; Davenport, Guy F; Fack, Veerle

    2018-05-31

    Core collections provide genebank curators and plant breeders a way to reduce size of their collections and populations, while minimizing impact on genetic diversity and allele frequency. Many methods have been proposed to generate core collections, often using distance metrics to quantify the similarity of two accessions, based on genetic marker data or phenotypic traits. Core Hunter is a multi-purpose core subset selection tool that uses local search algorithms to generate subsets relying on one or more metrics, including several distance metrics and allelic richness. In version 3 of Core Hunter (CH3) we have incorporated two new, improved methods for summarizing distances to quantify diversity or representativeness of the core collection. A comparison of CH3 and Core Hunter 2 (CH2) showed that these new metrics can be effectively optimized with less complex algorithms, as compared to those used in CH2. CH3 is more effective at maximizing the improved diversity metric than CH2, still ensures a high average and minimum distance, and is faster for large datasets. Using CH3, a simple stochastic hill-climber is able to find highly diverse core collections, and the more advanced parallel tempering algorithm further increases the quality of the core and further reduces variability across independent samples. We also evaluate the ability of CH3 to simultaneously maximize diversity, and either representativeness or allelic richness, and compare the results with those of the GDOpt and SimEli methods. CH3 can sample equally representative cores as GDOpt, which was specifically designed for this purpose, and is able to construct cores that are simultaneously more diverse, and either are more representative or have higher allelic richness, than those obtained by SimEli. In version 3, Core Hunter has been updated to include two new core subset selection metrics that construct cores for representativeness or diversity, with improved performance. It combines and outperforms the

  13. k-core covers and the core

    NARCIS (Netherlands)

    Sanchez-Rodriguez, E.; Borm, Peter; Estevez-Fernandez, A.; Fiestras-Janeiro, G.; Mosquera, M.A.

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  14. Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

    2015-10-15

    With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

  15. Thermal-Hydraulic Effects of Stud Shape and Size on the Safety Margin of Core Catcher System

    International Nuclear Information System (INIS)

    Song, Kyusang; Son, Hong Hyun; Jeong, Uiju; Kim, Sung Joong

    2015-01-01

    With the ERVC strategy, an additional system (core catcher system) to catch molten core penetrating the reactor pressure vessel (RPV) was proposed for advanced light water reactor. The newly engineered corium cooling system, that is, an ex-vessel core catcher system has been designed and adapted in some nuclear power plants such as VVER-1000, EPR, ESBWR, EU-APR1400 to mention a few. For example, Russia adopted a crucible-type core catcher for VVER-1000. On the other hand, a way to catch melt spreading is adopted by several countries, such as EPR in France, ESBWR in USA, ABWR in japan, and EU-APR1400 in Korea In Korea, the core catcher system has been designed and implemented for the European Advanced Power Reactor 1400 (EU-APR1400) to acquire a European license certificate. It is to confine molten materials in the reactor cavity while maintaining a coolable geometry in case that RPV failure occurs. The core catcher system consists of a carbon steel body, sacrificial material, protection material and engineered cooling channel. While installation of the studs is unavoidable, the studs tend to interfere in the smooth streamline of the core catcher channel. The distorted streamline could affect the overall thermal-hydraulic performance including two-phase heat transfer coefficient and critical heat flux (CHF) of the system. Thus, it is of importance to investigate the thermal-hydraulic effects of studs on the coolability, especially the CHF of the core catcher system. With aforementioned importance, pool boiling experiments were carried out with stud shape of, rectangular, cylinder, and elliptic and for stud sizes of 10, 15, 20, and 25 mm under the condition of atmospheric saturated water. A particular attention was focused on observing local vapor behavior around the studs and finding any hot spots, where the vapors are accumulated. The occurrence of the CHF is anticipated at the back side of the studs. The visual observation and CHF measurements indicate that the

  16. Intermittent degradation and schizotypy

    Directory of Open Access Journals (Sweden)

    Matthew W. Roché

    2015-06-01

    Full Text Available Intermittent degradation refers to transient detrimental disruptions in task performance. This phenomenon has been repeatedly observed in the performance data of patients with schizophrenia. Whether intermittent degradation is a feature of the liability for schizophrenia (i.e., schizotypy is an open question. Further, the specificity of intermittent degradation to schizotypy has yet to be investigated. To address these questions, 92 undergraduate participants completed a battery of self-report questionnaires assessing schizotypy and psychological state variables (e.g., anxiety, depression, and their reaction times were recorded as they did so. Intermittent degradation was defined as the number of times a subject’s reaction time for questionnaire items met or exceeded three standard deviations from his or her mean reaction time after controlling for each item’s information processing load. Intermittent degradation scores were correlated with questionnaire scores. Our results indicate that intermittent degradation is associated with total scores on measures of positive and disorganized schizotypy, but unrelated to total scores on measures of negative schizotypy and psychological state variables. Intermittent degradation is interpreted as potentially derivative of schizotypy and a candidate endophenotypic marker worthy of continued research.

  17. How do polymers degrade?

    Science.gov (United States)

    Lyu, Suping

    2011-03-01

    Materials derived from agricultural products such as cellulose, starch, polylactide, etc. are more sustainable and environmentally benign than those derived from petroleum. However, applications of these polymers are limited by their processing properties, chemical and thermal stabilities. For example, polyethylene terephthalate fabrics last for many years under normal use conditions, but polylactide fabrics cannot due to chemical degradation. There are two primary mechanisms through which these polymers degrade: via hydrolysis and via oxidation. Both of these two mechanisms are related to combined factors such as monomer chemistry, chain configuration, chain mobility, crystallinity, and permeation to water and oxygen, and product geometry. In this talk, we will discuss how these materials degrade and how the degradation depends on these factors under application conditions. Both experimental studies and mathematical modeling will be presented.

  18. Purex diluent degradation

    International Nuclear Information System (INIS)

    Tallent, O.K.; Mailen, J.C.; Pannell, K.D.

    1984-02-01

    The chemical degradation of normal paraffin hydrocarbon (NPH) diluents both in the pure state and mixed with 30% tributyl phosphate (TBP) was investigated in a series of experiments. The results show that degradation of NPH in the TBP-NPH-HNO 3 system is consistent with the active chemical agent being a radical-like nitrogen dioxide (NO 2 ) molecule, not HNO 3 as such. Spectrophotometric, gas chromatographic, mass spectrographic, and titrimetric methods were used to identify the degradation products, which included alkane nitro and nitrate compounds, alcohols, unsaturated alcohols, nitro alcohols, nitro alkenes, ketones, and carboxylic acids. The degradation rate was found to increase with increases in the HNO 3 concentration and the temperature. The rate was decreased by argon sparging to remove NO 2 and by the addition of butanol, which probably acts as a NO 2 scavenger. 13 references, 11 figures

  19. Bacteria and lignin degradation

    Institute of Scientific and Technical Information of China (English)

    Jing LI; Hongli YUAN; Jinshui YANG

    2009-01-01

    Lignin is both the most abundant aromatic (phenolic) polymer and the second most abundant raw material.It is degraded and modified by bacteria in the natural world,and bacteria seem to play a leading role in decomposing lignin in aquatic ecosystems.Lignin-degrading bacteria approach the polymer by mechanisms such as tunneling,erosion,and cavitation.With the advantages of immense environmental adaptability and biochemical versatility,bacteria deserve to be studied for their ligninolytic potential.

  20. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  1. Nuclear reactor core catcher

    International Nuclear Information System (INIS)

    1977-01-01

    A nuclear reactor core catcher is described for containing debris resulting from an accident causing core meltdown and which incorporates a method of cooling the debris by the circulation of a liquid coolant. (U.K.)

  2. Seismic core shroud

    International Nuclear Information System (INIS)

    Puri, A.; Mullooly, J.F.

    1981-01-01

    A core shroud is provided, comprising: a coolant boundary, following the shape of the core boundary, for channeling the coolant through the fuel assemblies; a cylindrical band positioned inside the core barrel and surrounding the coolant boundary; and support members extending from the coolant boundary to the band, for transferring load from the coolant boundary to the band. The shroud may be assembled in parts using automated welding techniques, and it may be adjusted to fit the reactor core easily

  3. Core Values | NREL

    Science.gov (United States)

    Core Values Core Values NREL's core values are rooted in a safe and supportive work environment guide our everyday actions and efforts: Safe and supportive work environment Respect for the rights physical and social environment Integrity Maintain the highest standard of ethics, honesty, and integrity

  4. Sidewall coring shell

    Energy Technology Data Exchange (ETDEWEB)

    Edelman, Ya A; Konstantinov, L P; Martyshin, A N

    1966-12-12

    A sidewall coring shell consists of a housing and a detachable core catcher. The core lifter is provided with projections, the ends of which are situated in another plane, along the longitudinal axis of the lifter. The chamber has corresponding projections.

  5. Rotary core drills

    Energy Technology Data Exchange (ETDEWEB)

    1967-11-30

    The design of a rotary core drill is described. Primary consideration is given to the following component parts of the drill: the inner and outer tube, the core bit, an adapter, and the core lifter. The adapter has the form of a downward-converging sleeve and is mounted to the lower end of the inner tube. The lifter, extending from the adapter, is split along each side so that it can be held open to permit movement of a core. It is possible to grip a core by allowing the lifter to assume a closed position.

  6. Drift Degradation Analysis

    International Nuclear Information System (INIS)

    D. Kicker

    2004-01-01

    Degradation of underground openings as a function of time is a natural and expected occurrence for any subsurface excavation. Over time, changes occur to both the stress condition and the strength of the rock mass due to several interacting factors. Once the factors contributing to degradation are characterized, the effects of drift degradation can typically be mitigated through appropriate design and maintenance of the ground support system. However, for the emplacement drifts of the geologic repository at Yucca Mountain, it is necessary to characterize drift degradation over a 10,000-year period, which is well beyond the functional period of the ground support system. This document provides an analysis of the amount of drift degradation anticipated in repository emplacement drifts for discrete events and time increments extending throughout the 10,000-year regulatory period for postclosure performance. This revision of the drift degradation analysis was developed to support the license application and fulfill specific agreement items between the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE). The earlier versions of ''Drift Degradation Analysis'' (BSC 2001 [DIRS 156304]) relied primarily on the DRKBA numerical code, which provides for a probabilistic key-block assessment based on realistic fracture patterns determined from field mapping in the Exploratory Studies Facility (ESF) at Yucca Mountain. A key block is defined as a critical block in the surrounding rock mass of an excavation, which is removable and oriented in an unsafe manner such that it is likely to move into an opening unless support is provided. However, the use of the DRKBA code to determine potential rockfall data at the repository horizon during the postclosure period has several limitations: (1) The DRKBA code cannot explicitly apply dynamic loads due to seismic ground motion. (2) The DRKBA code cannot explicitly apply loads due to thermal stress. (3) The DRKBA

  7. Drift Degradation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    D. Kicker

    2004-09-16

    Degradation of underground openings as a function of time is a natural and expected occurrence for any subsurface excavation. Over time, changes occur to both the stress condition and the strength of the rock mass due to several interacting factors. Once the factors contributing to degradation are characterized, the effects of drift degradation can typically be mitigated through appropriate design and maintenance of the ground support system. However, for the emplacement drifts of the geologic repository at Yucca Mountain, it is necessary to characterize drift degradation over a 10,000-year period, which is well beyond the functional period of the ground support system. This document provides an analysis of the amount of drift degradation anticipated in repository emplacement drifts for discrete events and time increments extending throughout the 10,000-year regulatory period for postclosure performance. This revision of the drift degradation analysis was developed to support the license application and fulfill specific agreement items between the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE). The earlier versions of ''Drift Degradation Analysis'' (BSC 2001 [DIRS 156304]) relied primarily on the DRKBA numerical code, which provides for a probabilistic key-block assessment based on realistic fracture patterns determined from field mapping in the Exploratory Studies Facility (ESF) at Yucca Mountain. A key block is defined as a critical block in the surrounding rock mass of an excavation, which is removable and oriented in an unsafe manner such that it is likely to move into an opening unless support is provided. However, the use of the DRKBA code to determine potential rockfall data at the repository horizon during the postclosure period has several limitations: (1) The DRKBA code cannot explicitly apply dynamic loads due to seismic ground motion. (2) The DRKBA code cannot explicitly apply loads due to thermal

  8. HYDRATE CORE DRILLING TESTS

    Energy Technology Data Exchange (ETDEWEB)

    John H. Cohen; Thomas E. Williams; Ali G. Kadaster; Bill V. Liddell

    2002-11-01

    The ''Methane Hydrate Production from Alaskan Permafrost'' project is a three-year endeavor being conducted by Maurer Technology Inc. (MTI), Noble, and Anadarko Petroleum, in partnership with the U.S. DOE National Energy Technology Laboratory (NETL). The project's goal is to build on previous and ongoing R&D in the area of onshore hydrate deposition. The project team plans to design and implement a program to safely and economically drill, core and produce gas from arctic hydrates. The current work scope includes drilling and coring one well on Anadarko leases in FY 2003 during the winter drilling season. A specially built on-site core analysis laboratory will be used to determine some of the physical characteristics of the hydrates and surrounding rock. Prior to going to the field, the project team designed and conducted a controlled series of coring tests for simulating coring of hydrate formations. A variety of equipment and procedures were tested and modified to develop a practical solution for this special application. This Topical Report summarizes these coring tests. A special facility was designed and installed at MTI's Drilling Research Center (DRC) in Houston and used to conduct coring tests. Equipment and procedures were tested by cutting cores from frozen mixtures of sand and water supported by casing and designed to simulate hydrate formations. Tests were conducted with chilled drilling fluids. Tests showed that frozen core can be washed out and reduced in size by the action of the drilling fluid. Washing of the core by the drilling fluid caused a reduction in core diameter, making core recovery very difficult (if not impossible). One successful solution was to drill the last 6 inches of core dry (without fluid circulation). These tests demonstrated that it will be difficult to capture core when drilling in permafrost or hydrates without implementing certain safeguards. Among the coring tests was a simulated hydrate

  9. Developments in polymer degradation - 7

    International Nuclear Information System (INIS)

    Grassie, N.

    1987-01-01

    A selection of topics which are representative of the continually expanding area of polymer degradation is presented. The aspects emphasised include the products of degradation of specific polymers, degradation by high energy radiation and mechanical forces, fire retardant studies and the special role of small radicals in degradation processes. (author)

  10. Motor degradation prediction methods

    Energy Technology Data Exchange (ETDEWEB)

    Arnold, J.R.; Kelly, J.F.; Delzingaro, M.J.

    1996-12-01

    Motor Operated Valve (MOV) squirrel cage AC motor rotors are susceptible to degradation under certain conditions. Premature failure can result due to high humidity/temperature environments, high running load conditions, extended periods at locked rotor conditions (i.e. > 15 seconds) or exceeding the motor`s duty cycle by frequent starts or multiple valve stroking. Exposure to high heat and moisture due to packing leaks, pressure seal ring leakage or other causes can significantly accelerate the degradation. ComEd and Liberty Technologies have worked together to provide and validate a non-intrusive method using motor power diagnostics to evaluate MOV rotor condition and predict failure. These techniques have provided a quick, low radiation dose method to evaluate inaccessible motors, identify degradation and allow scheduled replacement of motors prior to catastrophic failures.

  11. Endocytic collagen degradation

    DEFF Research Database (Denmark)

    Madsen, Daniel H.; Jürgensen, Henrik J.; Ingvarsen, Signe Ziir

    2012-01-01

    it crucially important to understand both the collagen synthesis and turnover mechanisms in this condition. Here we show that the endocytic collagen receptor, uPARAP/Endo180, is a major determinant in governing the balance between collagen deposition and degradation. Cirrhotic human livers displayed a marked...... up-regulation of uPARAP/Endo180 in activated fibroblasts and hepatic stellate cells located close to the collagen deposits. In a hepatic stellate cell line, uPARAP/Endo180 was shown to be active in, and required for, the uptake and intracellular degradation of collagen. To evaluate the functional...... groups of mice clearly revealed a fibrosis protective role of uPARAP/Endo180. This effect appeared to directly reflect the activity of the collagen receptor, since no compensatory events were noted when comparing the mRNA expression profiles of the two groups of mice in an array system focused on matrix-degrading...

  12. Degradation of fluorotelomer alcohols

    DEFF Research Database (Denmark)

    Ellis, David A; Martin, Jonathan W; De Silva, Amila O

    2004-01-01

    Human and animal tissues collected in urban and remote global locations contain persistent and bioaccumulative perfluorinated carboxylic acids (PFCAs). The source of PFCAs was previously unknown. Here we present smog chamber studies that indicate fluorotelomer alcohols (FTOHs) can degrade...... in the atmosphere to yield a homologous series of PFCAs. Atmospheric degradation of FTOHs is likely to contribute to the widespread dissemination of PFCAs. After their bioaccumulation potential is accounted for, the pattern of PFCAs yielded from FTOHs could account for the distinct contamination profile of PFCAs....... The significance of the gas-phase peroxy radical cross reactions that produce PFCAs has not been recognized previously. Such reactions are expected to occur during the atmospheric degradation of all polyfluorinated materials, necessitating a reexamination of the environmental fate and impact of this important...

  13. Motor degradation prediction methods

    International Nuclear Information System (INIS)

    Arnold, J.R.; Kelly, J.F.; Delzingaro, M.J.

    1996-01-01

    Motor Operated Valve (MOV) squirrel cage AC motor rotors are susceptible to degradation under certain conditions. Premature failure can result due to high humidity/temperature environments, high running load conditions, extended periods at locked rotor conditions (i.e. > 15 seconds) or exceeding the motor's duty cycle by frequent starts or multiple valve stroking. Exposure to high heat and moisture due to packing leaks, pressure seal ring leakage or other causes can significantly accelerate the degradation. ComEd and Liberty Technologies have worked together to provide and validate a non-intrusive method using motor power diagnostics to evaluate MOV rotor condition and predict failure. These techniques have provided a quick, low radiation dose method to evaluate inaccessible motors, identify degradation and allow scheduled replacement of motors prior to catastrophic failures

  14. Ecosystem degradation in India

    International Nuclear Information System (INIS)

    Sinha, B.N.

    1990-01-01

    Environmental and ecosystem studies have assumed greater relevance in the last decade of the twentieth century than even before. The urban settlements are becoming over-crowded and industries are increasingly polluting the air, water and sound in our larger metropolises. Degradation of different types of ecosystem are discussed in this book, Ecosystem Degradation in India. The book has been divided into seven chapters: Introduction, Coastal and Delta Ecosystem, River Basin Ecosystem, Mountain Ecosystem, Forest Ecosystem, Urban Ecosystem and the last chapter deals with the Environmental Problems and Planning. In the introduction the environmental and ecosystem degradation problems in India is highlighted as a whole while in other chapters mostly case studies by experts who know their respective terrain very intimately are included. The case study papers cover most part of India and deal with local problems, stretching from east coast to west coast and from Kashmir to Kanyakumari. (author)

  15. Antifoam degradation testing

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Newell, D. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River Ecology Lab. (SREL)

    2015-08-20

    This report describes the results of testing to quantify the degradation products resulting from the dilution and storage of Antifoam 747. Antifoam degradation is of concern to the Defense Waste Processing Facility (DWPF) due to flammable decomposition products in the vapor phase of the Chemical Process Cell vessels, as well as the collection of flammable and organic species in the offgas condensate. The discovery that hexamethyldisiloxane is formed from the antifoam decomposition was the basis for a Potential Inadequacy in the Safety Analysis declaration by the DWPF.

  16. Detection of pump degradation

    International Nuclear Information System (INIS)

    Casada, D.A.

    1994-01-01

    There are a variety of stressors that can affect the operation of centrifugal pumps. These can generally be classified as: Mechanical; Hydraulic; Tribological; Chemical; and Other (including those associated with the pump driver). Although these general stressors are active in essentially all centrifugal pumps, the stressor level and the extent of wear and degradation can vary greatly. Parameters that affect the extent of stressor activity are manifold. In order to assure the long-term operational readiness of a pump, it is important to both understand the nature and magnitude of the specific degradation mechanisms and to monitor the performance of the pump

  17. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  18. The core paradox.

    Science.gov (United States)

    Kennedy, G. C.; Higgins, G. H.

    1973-01-01

    Rebuttal of suggestions from various critics attempting to provide an escape from the seeming paradox originated by Higgins and Kennedy's (1971) proposed possibility that the liquid in the outer core was thermally stably stratified and that this stratification might prove a powerful inhibitor to circulation of the outer core fluid of the kind postulated for the generation of the earth's magnetic field. These suggestions are examined and shown to provide no reasonable escape from the core paradox.

  19. Drift Degradation Analysis

    International Nuclear Information System (INIS)

    G.H. Nieder-Westermann

    2005-01-01

    The outputs from the drift degradation analysis support scientific analyses, models, and design calculations, including the following: (1) Abstraction of Drift Seepage; (2) Seismic Consequence Abstraction; (3) Structural Stability of a Drip Shield Under Quasi-Static Pressure; and (4) Drip Shield Structural Response to Rock Fall. This report has been developed in accordance with ''Technical Work Plan for: Regulatory Integration Modeling of Drift Degradation, Waste Package and Drip Shield Vibratory Motion and Seismic Consequences'' (BSC 2004 [DIRS 171520]). The drift degradation analysis includes the development and validation of rockfall models that approximate phenomenon associated with various components of rock mass behavior anticipated within the repository horizon. Two drift degradation rockfall models have been developed: the rockfall model for nonlithophysal rock and the rockfall model for lithophysal rock. These models reflect the two distinct types of tuffaceous rock at Yucca Mountain. The output of this modeling and analysis activity documents the expected drift deterioration for drifts constructed in accordance with the repository layout configuration (BSC 2004 [DIRS 172801])

  20. Bacterial Degradation of Pesticides

    DEFF Research Database (Denmark)

    Knudsen, Berith Elkær

    could potentially improve bioremediation of BAM. An important prerequisite for bioaugmentation is the potential to produce the degrader strain at large quantities within reasonable time. The aim of manuscript II, was to optimize the growth medium for Aminobacter MSH1 and to elucidate optimal growth...

  1. Radiation degradation of silk

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, Kazushige; Kamiishi, Youichi [Textile Research Institute of Gunma, Kiryu, Gunma (Japan); Takeshita, Hidefumi; Yoshii, Fumio; Kume, Tamikazu [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    2001-03-01

    Silk fibroin powder was prepared from irradiated silk fibroin fiber by means of only physical treatment. Silk fibroin fiber irradiated with an accelerated electron beam in the dose range of 250 - 1000 kGy was pulverized by using a ball mill. Unirradiated silk fibroin fiber was not pulverized at all. But the more irradiation was increased, the more the conversion efficiency from fiber to powder was increased. The conversion efficiency of silk fibroin fiber irradiated 1000 kGy in oxygen was 94%. Silk fibroin powder shows remarkable solubility, which dissolved 57% into water of ambient temperature. It is a very interesting phenomenon that silk fibroin which did not treat with chemicals gets solubility only being pulverized. In order to study mechanism of solubilization of silk fibroin powder, amino acid component of soluble part of silk fibroin powder was analyzed. The more irradiation dose up, the more glycine or alanine degraded, but degradation fraction reached bounds about 50%. Other amino acids were degraded only 20% even at the maximum. To consider crystal construction of silk fibroin, it is suggested that irradiation on silk fibroin fiber selectively degrades glycine and alanine in amorphous region, which makes it possible to pulverize and to dissolve silk fibroin powder. (author)

  2. Nuclear reactor core flow baffling

    International Nuclear Information System (INIS)

    Berringer, R.T.

    1979-01-01

    A flow baffling arrangement is disclosed for the core of a nuclear reactor. A plurality of core formers are aligned with the grids of the core fuel assemblies such that the high pressure drop areas in the core are at the same elevations as the high pressure drop areas about the core periphery. The arrangement minimizes core bypass flow, maintains cooling of the structure surrounding the core, and allows the utilization of alternative beneficial components such as neutron reflectors positioned near the core

  3. Sediment Core Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — FUNCTION: Provides instrumentation and expertise for physical and geoacoustic characterization of marine sediments.DESCRIPTION: The multisensor core logger measures...

  4. Detection of pump degradation

    International Nuclear Information System (INIS)

    Greene, R.H.; Casada, D.A.; Ayers, C.W.

    1995-08-01

    This Phase II Nuclear Plant Aging Research study examines the methods of detecting pump degradation that are currently employed in domestic and overseas nuclear facilities. This report evaluates the criteria mandated by required pump testing at U.S. nuclear power plants and compares them to those features characteristic of state-of-the-art diagnostic programs and practices currently implemented by other major industries. Since the working condition of the pump driver is crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights recent developments in this technology. The routine collection and analysis of spectral data is superior to all other technologies in its ability to accurately detect numerous types and causes of pump degradation. Existing ASME Code testing criteria do not require the evaluation of pump vibration spectra but instead overall vibration amplitude. The mechanical information discernible from vibration amplitude analysis is limited, and several cases of pump failure were not detected in their early stages by vibration monitoring. Since spectral analysis can provide a wealth of pertinent information concerning the mechanical condition of rotating machinery, its incorporation into ASME testing criteria could merit a relaxation in the monthly-to-quarterly testing schedules that seek to verify and assure pump operability. Pump drivers are not included in the current battery of testing. Operational problems thought to be caused by pump degradation were found to be the result of motor degradation. Recent advances in nonintrusive monitoring techniques have made motor diagnostics a viable technology for assessing motor operability. Motor current/power analysis can detect rotor bar degradation and ascertain ranges of hydraulically unstable operation for a particular pump and motor set. The concept of using motor current or power fluctuations as an indicator of pump hydraulic load stability is presented

  5. Detection of pump degradation

    Energy Technology Data Exchange (ETDEWEB)

    Greene, R.H.; Casada, D.A.; Ayers, C.W. [and others

    1995-08-01

    This Phase II Nuclear Plant Aging Research study examines the methods of detecting pump degradation that are currently employed in domestic and overseas nuclear facilities. This report evaluates the criteria mandated by required pump testing at U.S. nuclear power plants and compares them to those features characteristic of state-of-the-art diagnostic programs and practices currently implemented by other major industries. Since the working condition of the pump driver is crucial to pump operability, a brief review of new applications of motor diagnostics is provided that highlights recent developments in this technology. The routine collection and analysis of spectral data is superior to all other technologies in its ability to accurately detect numerous types and causes of pump degradation. Existing ASME Code testing criteria do not require the evaluation of pump vibration spectra but instead overall vibration amplitude. The mechanical information discernible from vibration amplitude analysis is limited, and several cases of pump failure were not detected in their early stages by vibration monitoring. Since spectral analysis can provide a wealth of pertinent information concerning the mechanical condition of rotating machinery, its incorporation into ASME testing criteria could merit a relaxation in the monthly-to-quarterly testing schedules that seek to verify and assure pump operability. Pump drivers are not included in the current battery of testing. Operational problems thought to be caused by pump degradation were found to be the result of motor degradation. Recent advances in nonintrusive monitoring techniques have made motor diagnostics a viable technology for assessing motor operability. Motor current/power analysis can detect rotor bar degradation and ascertain ranges of hydraulically unstable operation for a particular pump and motor set. The concept of using motor current or power fluctuations as an indicator of pump hydraulic load stability is presented.

  6. In-core melt progression for the MAAP 4 codes

    International Nuclear Information System (INIS)

    Wu, C.-D.; Paik, Chan Y.; Henry, Robert E.; Ply, Martin G.

    2004-01-01

    The MAAP 4 core melt progression model contains provisions for the formation of a molten debris pool surrounded by a crust during late phase core degradation. A predominantly oxidic molten pool with a predominantly metallic lower crust may naturally develop through a combination of models for real material phase diagrams, mechanistic relocation, and rules to recognize extremely low porosity and the liquid fractions of adjacent highly degraded nodes. Pool size and shape thus becomes relatively independent of core nodalization (which only governs the coarseness of the crust location). An upper pool crust is mechanistically allowed during consideration of radiative and convective heat losses from the pool top surface to surrounding core nodes, the core barrel, and upper internals. Circulation within the pool causes mass and energy exchange between participating pool nodes, and determines the heat fluxes to the boundary crusts. Side and bottom node failure is predicted based on the time, temperature, and stress. Calculations demonstrate that this concept allows simulation of the degraded core geometry observed during the TMI-2 accident. (author)

  7. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    International Nuclear Information System (INIS)

    CUNNANE, J.

    2004-01-01

    Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i

  8. Can Psychiatric Rehabilitation Be Core to CORE?

    Science.gov (United States)

    Olney, Marjorie F.; Gill, Kenneth J.

    2016-01-01

    Purpose: In this article, we seek to determine whether psychiatric rehabilitation principles and practices have been more fully incorporated into the Council on Rehabilitation Education (CORE) standards, the extent to which they are covered in four rehabilitation counseling "foundations" textbooks, and how they are reflected in the…

  9. Liquid metal reactor core material HT9

    International Nuclear Information System (INIS)

    Kim, S. H.; Kuk, I. H.; Ryu, W. S. and others

    1998-03-01

    A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

  10. PWR core design calculations

    International Nuclear Information System (INIS)

    Trkov, A.; Ravnik, M.; Zeleznik, N.

    1992-01-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [sl

  11. Replaceable LMFBR core components

    International Nuclear Information System (INIS)

    Evans, E.A.; Cunningham, G.W.

    1976-01-01

    Much progress has been made in understanding material and component performance in the high temperature, fast neutron environment of the LMFBR. Current data have provided strong assurance that the initial core component lifetime objectives of FFTF and CRBR can be met. At the same time, this knowledge translates directly into the need for improved core designs that utilize improved materials and advanced fuels required to meet objectives of low doubling times and extended core component lifetimes. An industrial base for the manufacture of quality core components has been developed in the US, and all procurements for the first two core equivalents for FFTF will be completed this year. However, the problem of fabricating recycled plutonium while dramatically reducing fabrication costs, minimizing personnel exposure, and protecting public health and safety must be addressed

  12. Lunar Core and Tides

    Science.gov (United States)

    Williams, J. G.; Boggs, D. H.; Ratcliff, J. T.

    2004-01-01

    Variations in rotation and orientation of the Moon are sensitive to solid-body tidal dissipation, dissipation due to relative motion at the fluid-core/solid-mantle boundary, and tidal Love number k2 [1,2]. There is weaker sensitivity to flattening of the core-mantle boundary (CMB) [2,3,4] and fluid core moment of inertia [1]. Accurate Lunar Laser Ranging (LLR) measurements of the distance from observatories on the Earth to four retroreflector arrays on the Moon are sensitive to lunar rotation and orientation variations and tidal displacements. Past solutions using the LLR data have given results for dissipation due to solid-body tides and fluid core [1] plus Love number [1-5]. Detection of CMB flattening, which in the past has been marginal but improving [3,4,5], now seems significant. Direct detection of the core moment has not yet been achieved.

  13. Internal core tightener

    International Nuclear Information System (INIS)

    Brynsvold, G.V.; Snyder, H.J. Jr.

    1976-01-01

    An internal core tightener is disclosed which is a linear actuated (vertical actuation motion) expanding device utilizing a minimum of moving parts to perform the lateral tightening function. The key features are: (1) large contact areas to transmit loads during reactor operation; (2) actuation cam surfaces loaded only during clamping and unclamping operation; (3) separation of the parts and internal operation involved in the holding function from those involved in the actuation function; and (4) preloaded pads with compliant travel at each face of the hexagonal assembly at the two clamping planes to accommodate thermal expansion and irradiation induced swelling. The latter feature enables use of a ''fixed'' outer core boundary, and thus eliminates the uncertainty in gross core dimensions, and potential for rapid core reactivity changes as a result of core dimensional change. 5 claims, 12 drawing figures

  14. TALSPEAK Solvent Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Leigh R. Martin; Bruce J. Mincher

    2009-09-01

    Understanding the radiolytic degradation behavior of organic molecules involved in new or existing schemes for the recycle of used nuclear fuels is of significant interest for sustaining a closed nuclear fuel cycle. Here we have conducted several lines of investigation to begin understanding the effects of radiolysis on the aqueous phase of the TALSPEAK process for the separation of the trivalent lanthanides from the trivalent actinides. Using the 60-Co irradiator at the INL, we have begun to quantify the effects of radiation on the aqueous phase complexants used in this separation technique, and how this will affect the actinide lanthanide separation factor. In addition we have started to develop methodologies for stable product identification, a key element in determining the degradation pathways. We have also introduced a methodology to investigate the effects of alpha radiolysis that has previously received limited attention.

  15. Polyester-Based, Biodegradable Core-Multishell Nanocarriers for the Transport of Hydrophobic Drugs

    Directory of Open Access Journals (Sweden)

    Karolina A. Walker

    2016-05-01

    Full Text Available A water-soluble, core-multishell (CMS nanocarrier based on a new hyperbranched polyester core building block was synthesized and characterized towards drug transport and degradation of the nanocarrier. The hydrophobic drug dexamethasone was encapsulated and the enzyme-mediated biodegradability was investigated by NMR spectroscopy. The new CMS nanocarrier can transport one molecule of dexamethasone and degrades within five days at a skin temperature of 32 °C to biocompatible fragments.

  16. Rapidly Degradable Pyrotechnic System

    Science.gov (United States)

    2009-02-01

    material system (structural polymer and degradation agent ) for producing a high strength, non-corroding, highly inert, environmentally safe, extended...polymer sites in the active enzyme center differs dramatically between alkyl and aromatic polyesters. More specifically, as the degree of backbone...capped and centrifuged at 3,000 g. This procedure was repeated twice. To the remaining biomass pellet 15 mL of 1 mg/mL solution of N-ethyl-N- nitrosourea

  17. Radiation degradation of chitosan

    International Nuclear Information System (INIS)

    Norzita Yacob; Maznah Mahmud; Norhashidah Talip; Kamarudin Bahari; Kamaruddin Hashim; Khairul Zaman Dahlan

    2010-01-01

    In order to obtain an oligo chitosan, degradation of chitosan s were carried out in solid state and liquid state. The effects of an irradiation on the molecular weight and viscosity of the chitosan were investigated using Ubbelohde Capillary Viscometer and Brookfield Viscometer respectively. The molecular weight and viscosity of the chitosan s were decreased with an increase in the irradiation dose. In the presence of hydrogen peroxide, the molecular weight of chitosan can be further decreased. (author)

  18. Detection of pump degradation

    International Nuclear Information System (INIS)

    Casada, D.

    1994-01-01

    There are a variety of stressors that can affect the operation of centrifugal pumps. Although these general stressors are active in essentially all centrifugal pumps, the stressor level and the extent of wear and degradation can vary greatly. Parameters that affect the extent of stressor activity are manifold. In order to assure the long-term operational readiness of a pump, it is important to both understand the nature and magnitude of the specific degradation mechanisms and to monitor the performance of the pump. The most commonly applied method of monitoring the condition of not only pumps, but rotating machinery in general, is vibration analysis. Periodic or continuous spectral vibration analysis is a cornerstone of most pump monitoring programs. In the nuclear industry, non-spectral vibration monitoring of safety-related pumps is performed in accordance with the ASME code. Although vibration analysis has dominated the condition monitoring field for many years, there are other measures that have been historically used to help understand pump condition: advances in historically applied technologies and developing technologies offer improved monitoring capabilities. The capabilities of several technologies (including vibration analysis, dynamic pressure analysis, and motor power analysis) to detect the presence and magnitude of both stressors and resultant degradation are discussed

  19. Detection of pump degradation

    International Nuclear Information System (INIS)

    Casada, D.

    1995-01-01

    There are a variety of stressors that can affect the operation of centrifugal pumps. Although these general stressors are active in essentially all centrifugal pumps, the stressor level and the extent of wear and degradation can vary greatly. Parameters that affect the extent of stressor activity are manifold. In order to assure the long-term operational readiness of a pump, it is important to both understand the nature and magnitude of the specific degradation mechanisms and to monitor the performance of the pump. The most commonly applied method of monitoring the condition of not only pumps, but rotating machinery in general, is vibration analysis. Periodic or continuous special vibration analysis is a cornerstone of most pump monitoring programs. In the nuclear industry, non-spectral vibration monitoring of safety-related pumps is performed in accordance with the ASME code. Pump head and flow rate are also monitored, per code requirements. Although vibration analysis has dominated the condition monitoring field for many years, there are other measures that have been historically used to help understand pump condition; advances in historically applied technologies and developing technologies offer improved monitoring capabilities. The capabilities of several technologies (including vibration analysis, dynamic pressure analysis, and motor power analysis) to detect the presence and magnitude of both stressors and resultant degradation are discussed

  20. WEATHERABILITY OF ENHANCED DEGRADABLE PLASTICS

    Science.gov (United States)

    The main objective of this study was to assess the performance and the asociated variability of several selected enhanced degradable plastic materials under a variety of different exposure conditions. Other objectives were to identify the major products formed during degradation ...

  1. Interpretation of the results of the CORA-33 dry core BWR test

    International Nuclear Information System (INIS)

    Ott, L.J.; Hagen, S.

    1993-01-01

    All BWR degraded core experiments performed prior to CORA-33 were conducted under ''wet'' core degradation conditions for which water remains within the core and continuous steaming feeds metal/steam oxidation reactions on the in-core metallic surfaces. However, one dominant set of accident scenarios would occur with reduced metal oxidation under ''dry'' core degradation conditions and, prior to CORA-33, this set had been neglected experimentally. The CORA-33 experiment was designed specifically to address this dominant set of BWR ''dry'' core severe accident scenarios and to partially resolve phenomenological uncertainties concerning the behavior of relocating metallic melts draining into the lower regions of a ''dry'' BWR core. CORA-33 was conducted on October 1, 1992, in the CORA tests facility at KfK. Review of the CORA-33 data indicates that the test objectives were achieved; that is, core degradation occurred at a core heatup rate and a test section axial temperature profile that are prototypic of full-core nuclear power plant (NPP) simulations at ''dry'' core conditions. Simulations of the CORA-33 test at ORNL have required modification of existing control blade/canister materials interaction models to include the eutectic melting of the stainless steel/Zircaloy interaction products and the heat of mixing of stainless steel and Zircaloy. The timing and location of canister failure and melt intrusion into the fuel assembly appear to be adequately simulated by the ORNL models. This paper will present the results of the posttest analyses carried out at ORNL based upon the experimental data and the posttest examination of the test bundle at KfK. The implications of these results with respect to degraded core modeling and the associated safety issues are also discussed

  2. Earth's inner core: Innermost inner core or hemispherical variations?

    NARCIS (Netherlands)

    Lythgoe, K. H.; Deuss, A.|info:eu-repo/dai/nl/412396610; Rudge, J. F.; Neufeld, J. A.

    2014-01-01

    The structure of Earth's deep inner core has important implications for core evolution, since it is thought to be related to the early stages of core formation. Previous studies have suggested that there exists an innermost inner core with distinct anisotropy relative to the rest of the inner core.

  3. Dependence of Core and Extended Flux on Core Dominance ...

    Indian Academy of Sciences (India)

    Abstract. Based on two extragalactic radio source samples, the core dominance parameter is calculated, and the correlations between the core/extended flux density and core dominance parameter are investi- gated. When the core dominance parameter is lower than unity, it is linearly correlated with the core flux density, ...

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

    OpenAIRE

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

    2010-01-01

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

  5. Windscale pile core surveys

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  6. Core shroud corner joints

    Science.gov (United States)

    Gilmore, Charles B.; Forsyth, David R.

    2013-09-10

    A core shroud is provided, which includes a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each unitary corner comprises a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies comprises a plurality of the unitary corners disposed side-by-side in an alternating opposing relationship. A plurality of the subassemblies can be combined to form a quarter perimeter segment of the core shroud. Four quarter perimeter segments join together to form the core shroud.

  7. IGCSE core mathematics

    CERN Document Server

    Wall, Terry

    2013-01-01

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

  8. On-line generation of core monitoring power distribution in the SCOMS couppled with core design code

    International Nuclear Information System (INIS)

    Lee, K. B.; Kim, K. K.; In, W. K.; Ji, S. K.; Jang, M. H.

    2002-01-01

    The paper provides the description of the methodology and main program module of power distribution calculation of SCOMS(SMART COre Monitoring System). The simulation results of the SMART core using the developed SCOMS are included. The planar radial peaking factor(Fxy) is relatively high in SMART core because control banks are inserted to the core at normal operation. If the conventional core monitoring method is adapted to SMART, highly skewed planar radial peaking factor Fxy yields an excessive conservatism and reduces the operation margin. In addition to this, the error of the core monitoring would be enlarged and thus operating margin would be degraded, because it is impossible to precalculate the core monitoring constants for all the control banks configurations taking into account the operation history in the design stage. To get rid of these drawbacks in the conventional power distribution calculation methodology, new methodology to calculate the three dimensional power distribution is developed. Core monitoring constants are calculated with the core design code (MASTER) which is on-line coupled with SCOMS. Three dimensional (3D) power distribution and the several peaking factors are calculated using the in-core detector signals and core monitoring constant provided at real time. Developed methodology is applied to the SMART core and the various core states are simulated. Based on the simulation results, it is founded that the three dimensional peaking factor to calculate the Linear Power Density and the pseudo hot-pin axial power distribution to calculate the Departure Nucleate Boiling Ratio show the more conservative values than those of the best-estimated core design code, and SCOMS adapted developed methodology can secures the more operation margin than the conventional methodology

  9. Tailored Core Shell Cathode Powders for Solid Oxide Fuel Cells

    Energy Technology Data Exchange (ETDEWEB)

    Swartz, Scott [NexTech Materials, Ltd.,Lewis Center, OH (United States)

    2015-03-23

    In this Phase I SBIR project, a “core-shell” composite cathode approach was evaluated for improving SOFC performance and reducing degradation of lanthanum strontium cobalt ferrite (LSCF) cathode materials, following previous successful demonstrations of infiltration approaches for achieving the same goals. The intent was to establish core-shell cathode powders that enabled high performance to be obtained with “drop-in” process capability for SOFC manufacturing (i.e., rather than adding an infiltration step to the SOFC manufacturing process). Milling, precipitation and hetero-coagulation methods were evaluated for making core-shell composite cathode powders comprised of coarse LSCF “core” particles and nanoscale “shell” particles of lanthanum strontium manganite (LSM) or praseodymium strontium manganite (PSM). Precipitation and hetero-coagulation methods were successful for obtaining the targeted core-shell morphology, although perfect coverage of the LSCF core particles by the LSM and PSM particles was not obtained. Electrochemical characterization of core-shell cathode powders and conventional (baseline) cathode powders was performed via electrochemical impedance spectroscopy (EIS) half-cell measurements and single-cell SOFC testing. Reliable EIS testing methods were established, which enabled comparative area-specific resistance measurements to be obtained. A single-cell SOFC testing approach also was established that enabled cathode resistance to be separated from overall cell resistance, and for cathode degradation to be separated from overall cell degradation. The results of these EIS and SOFC tests conclusively determined that the core-shell cathode powders resulted in significant lowering of performance, compared to the baseline cathodes. Based on the results of this project, it was concluded that the core-shell cathode approach did not warrant further investigation.

  10. Degradation of AF1Q by chaperone-mediated autophagy

    International Nuclear Information System (INIS)

    Li, Peng; Ji, Min; Lu, Fei; Zhang, Jingru; Li, Huanjie; Cui, Taixing; Li Wang, Xing; Tang, Dongqi; Ji, Chunyan

    2014-01-01

    AF1Q, a mixed lineage leukemia gene fusion partner, is identified as a poor prognostic biomarker for pediatric acute myeloid leukemia (AML), adult AML with normal cytogenetic and adult myelodysplastic syndrome. AF1Q is highly regulated during hematopoietic progenitor differentiation and development but its regulatory mechanism has not been defined clearly. In the present study, we used pharmacological and genetic approaches to influence chaperone-mediated autophagy (CMA) and explored the degradation mechanism of AF1Q. Pharmacological inhibitors of lysosomal degradation, such as chloroquine, increased AF1Q levels, whereas activators of CMA, including 6-aminonicotinamide and nutrient starvation, decreased AF1Q levels. AF1Q interacts with HSPA8 and LAMP-2A, which are core components of the CMA machinery. Knockdown of HSPA8 or LAMP-2A increased AF1Q protein levels, whereas overexpression showed the opposite effect. Using an amino acid deletion AF1Q mutation plasmid, we identified that AF1Q had a KFERQ-like motif which was recognized by HSPA8 for CMA-dependent proteolysis. In conclusion, we demonstrate for the first time that AF1Q can be degraded in lysosomes by CMA. - Highlights: • Chaperone-mediated autophagy (CMA) is involved in the degradation of AF1Q. • Macroautophagy does not contribute to the AF1Q degradation. • AF1Q has a KFERQ-like motif that is recognized by CMA core components

  11. Degradation of AF1Q by chaperone-mediated autophagy

    Energy Technology Data Exchange (ETDEWEB)

    Li, Peng; Ji, Min; Lu, Fei; Zhang, Jingru [Department of Hematology, Key Laboratory of Cardiovascular Remodeling and Function Research, Qilu Hospital, Shandong University, Jinan 250012 (China); Li, Huanjie; Cui, Taixing; Li Wang, Xing [Research Center for Cell Therapy, Key Laboratory of Cardiovascular Remodeling and Function Research, Qilu Hospital, Shandong University, Jinan 250012 (China); Tang, Dongqi, E-mail: tangdq@sdu.edu.cn [Research Center for Cell Therapy, Key Laboratory of Cardiovascular Remodeling and Function Research, Qilu Hospital, Shandong University, Jinan 250012 (China); Center for Stem Cell and Regenerative Medicine, The Second Hospital of Shandong University, Jinan 250033 (China); Ji, Chunyan, E-mail: jichunyan@sdu.edu.cn [Department of Hematology, Key Laboratory of Cardiovascular Remodeling and Function Research, Qilu Hospital, Shandong University, Jinan 250012 (China)

    2014-09-10

    AF1Q, a mixed lineage leukemia gene fusion partner, is identified as a poor prognostic biomarker for pediatric acute myeloid leukemia (AML), adult AML with normal cytogenetic and adult myelodysplastic syndrome. AF1Q is highly regulated during hematopoietic progenitor differentiation and development but its regulatory mechanism has not been defined clearly. In the present study, we used pharmacological and genetic approaches to influence chaperone-mediated autophagy (CMA) and explored the degradation mechanism of AF1Q. Pharmacological inhibitors of lysosomal degradation, such as chloroquine, increased AF1Q levels, whereas activators of CMA, including 6-aminonicotinamide and nutrient starvation, decreased AF1Q levels. AF1Q interacts with HSPA8 and LAMP-2A, which are core components of the CMA machinery. Knockdown of HSPA8 or LAMP-2A increased AF1Q protein levels, whereas overexpression showed the opposite effect. Using an amino acid deletion AF1Q mutation plasmid, we identified that AF1Q had a KFERQ-like motif which was recognized by HSPA8 for CMA-dependent proteolysis. In conclusion, we demonstrate for the first time that AF1Q can be degraded in lysosomes by CMA. - Highlights: • Chaperone-mediated autophagy (CMA) is involved in the degradation of AF1Q. • Macroautophagy does not contribute to the AF1Q degradation. • AF1Q has a KFERQ-like motif that is recognized by CMA core components.

  12. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Diaz, N.J.; Dugan, E.T.

    1983-01-01

    A heterogeneous gas core nuclear reactor is disclosed comprising a core barrel provided interiorly with an array of moderator-containing tubes and being otherwise filled with a fissile and/or fertile gaseous fuel medium. The fuel medium may be flowed through the chamber and through an external circuit in which heat is extracted. The moderator may be a fluid which is flowed through the tubes and through an external circuit in which heat is extracted. The moderator may be a solid which may be cooled by a fluid flowing within the tubes and through an external heat extraction circuit. The core barrel is surrounded by moderator/coolant material. Fissionable blanket material may be disposed inwardly or outwardly of the core barrel

  13. iPSC Core

    Data.gov (United States)

    Federal Laboratory Consortium — The induced Pluripotent Stem Cells (iPSC) Core was created in 2011 to accelerate stem cell research in the NHLBI by providing investigators consultation, technical...

  14. Core Flight Software

    Data.gov (United States)

    National Aeronautics and Space Administration — The AES Core Flight Software (CFS) project purpose is to analyze applicability, and evolve and extend the reusability of the CFS system originally developed by...

  15. Statistical modeling for degradation data

    CERN Document Server

    Lio, Yuhlong; Ng, Hon; Tsai, Tzong-Ru

    2017-01-01

    This book focuses on the statistical aspects of the analysis of degradation data. In recent years, degradation data analysis has come to play an increasingly important role in different disciplines such as reliability, public health sciences, and finance. For example, information on products’ reliability can be obtained by analyzing degradation data. In addition, statistical modeling and inference techniques have been developed on the basis of different degradation measures. The book brings together experts engaged in statistical modeling and inference, presenting and discussing important recent advances in degradation data analysis and related applications. The topics covered are timely and have considerable potential to impact both statistics and reliability engineering.

  16. Restraint system for core elements of a reactor core

    International Nuclear Information System (INIS)

    Class, G.

    1975-01-01

    In a nuclear reactor, a core element bundle formed of a plurality of side-by-side arranged core elements is surrounded by restraining elements that exert a radially inwardly directly restraining force generating friction forces between the core elements in a restraining plane that is transverse to the core element axes. The adjoining core elements are in rolling contact with one another in the restraining plane by virtue of rolling-type bearing elements supported in the core elements. (Official Gazette)

  17. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector

    Directory of Open Access Journals (Sweden)

    Saas Laurent

    2017-01-01

    Full Text Available In the context of the simulation of the Severe Accidents (SA in Light Water Reactors (LWR, we are interested on the in-core corium pool propagation transient in order to evaluate the corium relocation in the vessel lower head. The goal is to characterize the corium and debris flows from the core to accurately evaluate the corium pool propagation transient in the lower head and so the associated risk of vessel failure. In the case of LWR with heavy reflector, to evaluate the corium relocation into the lower head, we have to study the risk associated with focusing effect and the possibility to stabilize laterally the corium in core with a flooded down-comer. It is necessary to characterize the core degradation and the stratification of the corium pool that is formed in core. We assume that the core degradation until the corium pool formation and the corium pool propagation could be modeled separately. In this document, we present a simplified geometrical model (0D model for the in-core corium propagation transient. A degraded core with a formed corium pool is used as an initial state. This state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate…. During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4.

  18. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  19. Development tendencies of moulding and core sands

    Directory of Open Access Journals (Sweden)

    Stanislaw M. Dobosz1

    2011-11-01

    Full Text Available Further development of the technology for making moulding and core sands will be strictly limited by tough requirements due to protection of the natural environment. These tendencies are becoming more and more tense, so that we will reach a point when even processes, that from technological point of view fulfill high requirements of the foundry industry, must be replaced by more ecologically-friendly solutions. Hence, technologies using synthetic resins as binding materials will be limited. This paper presents some predictable development tendencies of moulding and core sands. The increasing role of inorganic substances will be noticed, including silicate binders with significantly improved properties, such as improved knock-out property or higher reclamation strength. Other interesting solutions might also be moulding sands bonded by geo-polymers and phosphate binders or salts and also binders based on degradable biopolymers. These tendencies and the usefulness of these binders are put forward in this paper.

  20. Radiation degradation of cellulose

    International Nuclear Information System (INIS)

    Leonhardt, J.; Arnold, G.; Baer, M.; Langguth, H.; Gey, M.; Huebert, S.

    1985-01-01

    The application of straw and other cellulose polymers as feedstuff for ruminants is limited by its low digestibility. During recent decades it was attempted to increase the digestibility of straw by several chemical and physical methods. In this work some results of the degradation of gamma and electron treated wheat straw are reported. Complex methods of treatment are taken into consideration. In vitro-experiments with radiation treated straw show that the digestibility can be increased from 20% up to about 80%. A high pressure liquid chromatography method was used to analyze the hydrolysates. The contents of certain species of carbohydrates in the hydrolysates in dependence on the applied dose are given. (author)

  1. Chemical degradation of pentachlorophenol

    International Nuclear Information System (INIS)

    Shukla, S.S.; Shukla, A.; Chandrasekharaiah, M.S.

    1992-01-01

    Industry produces a large volume of hazardous wastes containing pentachlorophenol, a U.S. EPA priority hazardous organic material. The environmentally safe disposal of these PCP-contaminated wastes is a serious problem for the waste management authorities as the current treatment processes are unsatisfactory. In this paper, the results of a feasibility study of chemical degradation and/or solidification methods for PCP-containing wastes. The photochemical decomposition of the PCP in a microemulsion or in micellar media obtained with the help of SDS or CTAB show the greatest promise

  2. Radiation degradation of polymethacrylamide

    International Nuclear Information System (INIS)

    O'Connor, D.J.

    1984-01-01

    The effects of radiation on polymers have been studied for many years. When polymers are subjected to ultraviolet light or ionizing radiation, chain scission and crosslinking are possible. The radiation degradations of several methacrylate type polymers were investigated. The primary polymer studied was polymethacrylamide (PMAAm). Ultraviolet irradiated PMAAm yielded a five line ESR spectrum with 22 gauss splitting which is believed to arise from a polymeric radical ending with a methacrylamide unit. The results obtained indicate that polymethacrylamide is a polymer which undergoes main chain cleavage upon irradiation. As such this polymer may have potential applicability as a positive resist for fabrication of microelectronic devices

  3. FBR type reactor core

    International Nuclear Information System (INIS)

    Tamiya, Tadashi; Kawashima, Katsuyuki; Fujimura, Koji; Murakami, Tomoko.

    1995-01-01

    Neutron reflectors are disposed at the periphery of a reactor core fuel region and a blanket region, and a neutron shielding region is disposed at the periphery of them. The neutron reflector has a hollow duct structure having a sealed upper portion, a lower portion opened to cooling water, in which a gas and coolants separately sealed in the inside thereof. A driving pressure of a primary recycling pump is lowered upon reduction of coolant flow rate, then the liquid level of coolants in the neutron reflector is lowered due to imbalance between the driving pressure and a gas pressure, so that coolants having an effect as a reflector are eliminated from the outer circumference of the reactor core. Therefore, the amount of neutrons leaking from the reactor core is increased, and negative reactivity is charged to the reactor core. The negative reactivity of the neutron reflector is made greater than a power compensation reactivity. Since this enables reactor scram by using an inherent performance of the reactor core, the reactor core safety of an LMFBR-type reactor can be improved. (I.N.)

  4. The earths innermost core

    International Nuclear Information System (INIS)

    Nanda, J.N.

    1989-01-01

    A new earth model is advanced with a solid innermost core at the centre of the Earth where elements heavier than iron, over and above what can be retained in solution in the iron core, are collected. The innermost core is separated from the solid iron-nickel core by a shell of liquid copper. The innermost core has a natural vibration measured on the earth's surface as the long period 26 seconds microseisms. The earth was formed initially as a liquid sphere with a relatively thin solid crust above the Byerly discontinuity. The trace elements that entered the innermost core amounted to only 0.925 ppm of the molten mass. Gravitational differentiation must have led to the separation of an explosive thickness of pure 235 U causing a fission explosion that could expel beyond the Roche limit a crustal scab which would form the centre piece of the moon. A reservoir of helium floats on the liquid copper. A small proportion of helium-3, a relic of the ancient fission explosion present there will spell the exciting magnetic field. The field is stable for thousands of years because of the presence of large quantity of helium-4 which accounts for most of the gaseous collisions that will not disturb the atomic spin of helium-3 atoms. This field is prone to sudden reversals after long periods of stability. (author). 14 refs

  5. Sleeve type repair of degraded nuclear steam generator tubes

    International Nuclear Information System (INIS)

    Ayres, P.S.; Stark, L.E.; Feldstein, J.G.; Fu, T.

    1986-01-01

    A sealable sleeve is described for insertion into the repair of a degraded tube which consists of: a hollow core inner member of the same material as the degraded tube; a thinner outer member of substantially pure nickel and resistant to corrosive attack, the outer member being metallurgically bonded with the inner member; an expanded portion of the sleeve at one end for positioning in the tube within a tube sheet; a multiplicity of grooves formed in and adjacent to the other end of the sleeve which extends into the free-standing portion of the tube beyond the tube sheet, and a noble metal braze material contained in the grooves

  6. Materials Degradation and Detection (MD2): Deep Dive Final Report

    Energy Technology Data Exchange (ETDEWEB)

    McCloy, John S.; Montgomery, Robert O.; Ramuhalli, Pradeep; Meyer, Ryan M.; Hu, Shenyang Y.; Li, Yulan; Henager, Charles H.; Johnson, Bradley R.

    2013-02-01

    An effort is underway at Pacific Northwest National Laboratory (PNNL) to develop a fundamental and general framework to foster the science and technology needed to support real-time monitoring of early degradation in materials used in the production of nuclear power. The development of such a capability would represent a timely solution to the mounting issues operators face with materials degradation in nuclear power plants. The envisioned framework consists of three primary and interconnected “thrust” areas including 1) microstructural science, 2) behavior assessment, and 3) monitoring and predictive capabilities. A brief state-of-the-art assessment for each of these core technology areas is discussed in the paper.

  7. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

  8. The role of fission product in whole core accidents - research in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Dietrich, L W [Argonne National Laboratory, Division of Reactor Analysis and Safety, Argonne, IL (United States); Jackson, J F [Los Alamos Scientific Laboratory, Q Division - Energy, Los Alamos, NM (United States)

    1977-07-01

    Safety of nuclear reactors has been a central concern of the nuclear energy industry from the very beginning. This concern, and the resultant excellence of design, fabrication, and operation, aided by extensive engineered safety features, has given nuclear energy its superior record of protection of the environment and of the public health and safety. With respect to the fast reactor, it was recognized early in the programme that there exists a theoretical possibility of a core compaction leading to significant energy release. Early analysis of this problem employed a number of conservative assumptions in attempting to bound the energy release. As reactors have grown in size, the suitability of such bounding calculations has diminished, and research into hypothetical accident analysis has emphasized a more mechanistic approach. In the USA, much effort has been directed towards modeling and computer code development aimed at following the course of an accident from its initiation to its ultimate conclusion with a stable, permanently subcritical, coolable core geometry, along with considerations of post-accident heat removal and radiological consequence assessment. Throughout this effort, the potential role of fission products has been recognized and account taken of the effects of fission products in determining accident progression. It is important to recognize that reactor safety is a very diverse topic, requiring consideration of a number of factors. While the major questions of public risk appear to be related to the hypothetical core disruptive accident (HCDA), it is necessary that the probability of having such an accident be extremely low In order that acceptable public risk be demonstrated. Such a demonstration requires sound engineering design and Implementation, with high standards of reliability, inspectability, maintainability, and operation, along with the requisite quality control and assurance. Tile current approach, typified by that taken by the

  9. Degradable polyphosphazene/poly(alpha-hydroxyester) blends: degradation studies.

    Science.gov (United States)

    Ambrosio, Archel M A; Allcock, Harry R; Katti, Dhirendra S; Laurencin, Cato T

    2002-04-01

    Biomaterials based on the polymers of lactic acid and glycolic acid and their copolymers are used or studied extensively as implantable devices for drug delivery, tissue engineering and other biomedical applications. Although these polymers have shown good biocompatibility, concerns have been raised regarding their acidic degradation products, which have important implications for long-term implantable systems. Therefore, we have designed a novel biodegradable polyphosphazene/poly(alpha-hydroxyester) blend whose degradation products are less acidic than those of the poly(alpha-hydroxyester) alone. In this study, the degradation characteristics of a blend of poly(lactide-co-glycolide) (50:50 PLAGA) and poly[(50% ethyl glycinato)(50% p-methylphenoxy) phosphazene] (PPHOS-EG50) were qualitatively and quantitatively determined with comparisons made to the parent polymers. Circular matrices (14mm diameter) of the PLAGA, PPHOS-EG50 and PLAGA-PPHOS-EG50 blend were degraded in non-buffered solutions (pH 7.4). The degraded polymers were characterized for percentage mass loss and molecular weight and the degradation medium was characterized for acid released in non-buffered solutions. The amounts of neutralizing base necessary to bring about neutral pH were measured for each polymer or polymer blend during degradation. The poly(phosphazene)/poly(lactide-co-glycolide) blend required significantly less neutralizing base in order to bring about neutral solution pH during the degradation period studied. The results indicated that the blend degraded at a rate intermediate to that of the parent polymers and that the degradation products of the polyphosphazene neutralized the acidic degradation products of PLAGA. Thus, results from these in vitro degradation studies suggest that the PLAGA-PPHOS-EG50 blend may provide a viable improvement to biomaterials based on acid-releasing organic polymers.

  10. Reactor core control device

    International Nuclear Information System (INIS)

    Sano, Hiroki

    1998-01-01

    The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

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

    International Nuclear Information System (INIS)

    Platus, D.L.

    1976-01-01

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

  12. Sensitivity analysis using DECOMP and METOXA subroutines of the MAAP code in modelling core concrete interaction phenomena and post test calculations for ACE-MCCI experiment L-5

    International Nuclear Information System (INIS)

    Passalacqua, R.A.

    1991-01-01

    A parametric analysis approach was chosen in order to study core-concrete interaction phenomena. The analysis was performed using a stand-alone version of the MAAP-DECOMP model (DOE version). This analysis covered only those parameters known to have the largest effect on thermohydraulics and fission product aerosol release. Even though the main purpose of the effort was model validation, it eventually resulted in a better understanding of the core-concrete interaction physics and to a more correct interpretation of the ACE-MCCI L5 experimental data. Unusual low heat transfer fluxes from the debris pool to the cavity (corium surrounding volume) were modeled in order to have a good benchmark with the experimental data. Therefore, higher debris pool temperatures were predicted. In case of water flooding, as a consequence of the critical heat flux through the upper crust and the increase of the crust thickness, resulting high debris pool temperatures cause an increase in the concrete ablation rate in the short term. DECOMP model predicts a quick increase of the crust thickness and as a result, causes the quenching of the molten mass. However, especially for fast transient, phenomena of crust bridge formation can occur. Thus, the upward directed heat flux is minimized and the concrete erosion rate remains conspicuous also in the long term. The model validation is based, in these calculations, on post-test predictions using the MCCI L5 test data: these data are derived from results of the 'Molten Core Concrete Interaction' (MCCI) experiments, which in turn are part of the larger Advanced Containment Experiment (ACE) program. Other calculations were also performed for the new proposed MACE (Melt Debris Attack and Coolability) experiments simulating the water flooding of the cavity. Those calculations are preliminarily compared with the recent MACE scoping test results. (author) 4 tabs., 59 figs., 5 refs

  13. Radiation degradation of cellulose

    International Nuclear Information System (INIS)

    Leonhardt, J.W.; Arnold, G.; Baer, M.; Gey, M.; Hubert, S.; Langguth, H.

    1984-01-01

    The application of straw and other cellulose polymers as feedstuff for ruminants is limited by its low digestibility. During recent decades it was attempted to increase the digestibility of straw by several chemical and physical methods. In this work some results of the degradation of gamma and electron treated wheat straw are reported. Complex methods of treatment (e.g. radiation influence and influence of lyes) are taken into consideration. In vitro-experiments with radiation treated straw show that the digestibility can be increased from 20% up to about 80%. A high pressure liquid chromatography method was used to analyze the hydrolysates. The contents of certain species of carbohydrates in the hydrolysates in dependence on the applied dose are given

  14. Exploring bacterial lignin degradation.

    Science.gov (United States)

    Brown, Margaret E; Chang, Michelle C Y

    2014-04-01

    Plant biomass represents a renewable carbon feedstock that could potentially be used to replace a significant level of petroleum-derived chemicals. One major challenge in its utilization is that the majority of this carbon is trapped in the recalcitrant structural polymers of the plant cell wall. Deconstruction of lignin is a key step in the processing of biomass to useful monomers but remains challenging. Microbial systems can provide molecular information on lignin depolymerization as they have evolved to break lignin down using metalloenzyme-dependent radical pathways. Both fungi and bacteria have been observed to metabolize lignin; however, their differential reactivity with this substrate indicates that they may utilize different chemical strategies for its breakdown. This review will discuss recent advances in studying bacterial lignin degradation as an approach to exploring greater diversity in the environment. Copyright © 2013 Elsevier Ltd. All rights reserved.

  15. Soil degradation in Pakistan

    International Nuclear Information System (INIS)

    Khan, M.R.

    2005-01-01

    This paper diagnoses the issues involved behind the current state, usage, interactions and linkages in the soils in Pakistan. The condition of soils is deteriorating due to developmental and environmental factors such as soil degradation, water pollution, fauna degeneration etc. Issues, problems and constraints faced in the management and usage of soils are diagnosed at different levels in the ecosystems predominant in Pakistan. The research questions propose effective solutions, types of instruments, methods or processes to resolve the issues within the various areas or ecosystems in the most sustainable and effective manner [23]. Biological solutions and methods can be applied at the sub-system level by private individuals or communities at a lower cost, and at a more localized level than engineering methods. Engineering methods may be suited for interventions at a system level rather than at a sub-system level; but even at this level they will be complementary with biological methods. (author)

  16. The PWR cores management

    International Nuclear Information System (INIS)

    Barral, J.C.; Rippert, D.; Johner, J.

    2000-01-01

    During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

  17. Molten core retention assembly

    International Nuclear Information System (INIS)

    Lampe, R.F.

    1976-01-01

    Molten fuel produced in a core overheating accident is caught by a molten core retention assembly consisting of a horizontal baffle plate having a plurality of openings therein, heat exchange tubes having flow holes near the top thereof mounted in the openings, and a cylindrical imperforate baffle attached to the plate and surrounding the tubes. The baffle assembly is supported from the core support plate of the reactor by a plurality of hanger rods which are welded to radial beams passing under the baffle plate and intermittently welded thereto. Preferably the upper end of the cylindrical baffle terminates in an outwardly facing lip to which are welded a plurality of bearings having slots therein adapted to accept the hanger rods

  18. Core status computing system

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki.

    1982-01-01

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

  19. Degraded Crater Rim

    Science.gov (United States)

    2002-01-01

    (Released 3 May 2002) The Science The eastern rim of this unnamed crater in Southern Arabia Terra is very degraded (beaten up). This indicates that this crater is very ancient and has been subjected to erosion and subsequent bombardment from other impactors such as asteroids and comets. One of these later (younger) craters is seen in the upper right of this image superimposed upon the older crater rim material. Note that this smaller younger crater rim is sharper and more intact than the older crater rim. This region is also mantled with a blanket of dust. This dust mantle causes the underlying topography to take on a more subdued appearance. The Story When you think of Arabia, you probably think of hot deserts and a lot of profitable oil reserves. On Mars, however, Southern Arabia Terra is a cold place of cratered terrain. This almost frothy-looking image is the badly battered edge of an ancient crater, which has suffered both erosion and bombardment from asteroids, comets, or other impacting bodies over the long course of its existence. A blanket of dust has also settled over the region, which gives the otherwise rugged landscape a soft and more subdued appearance. The small, round crater (upper left) seems almost gemlike in its setting against the larger crater ring. But this companionship is no easy romance. Whatever formed the small crater clearly whammed into the larger crater rim at some point, obliterating part of its edge. You can tell the small crater was formed after the first and more devastating impact, because it is laid over the other larger crater. How much younger is the small one? Well, its rim is also much sharper and more intact, which gives a sense that it is probably far more youthful than the very degraded, ancient crater.

  20. Superconducting tin core fiber

    International Nuclear Information System (INIS)

    Homa, Daniel; Liang, Yongxuan; Hill, Cary; Kaur, Gurbinder; Pickrell, Gary

    2015-01-01

    In this study, we demonstrated superconductivity in a fiber with a tin core and fused silica cladding. The fibers were fabricated via a modified melt-draw technique and maintained core diameters ranging from 50-300 microns and overall diameters of 125-800 microns. Superconductivity of this fiber design was validated via the traditional four-probe test method in a bath of liquid helium at temperatures on the order of 3.8 K. The synthesis route and fiber design are perquisites to ongoing research dedicated all-fiber optoelectronics and the relationships between superconductivity and the material structures, as well as corresponding fabrication techniques. (orig.)

  1. LMFBR core design analysis

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  2. Nuclear core catchers

    International Nuclear Information System (INIS)

    Golden, M.P.; Tilbrook, R.W.; Heylmun, N.F.

    1976-01-01

    A receptacle is described for taking the molten fragments of a nuclear reactor during a reactor core fusion accident. The receptacle is placed under the reactor. It includes at least one receptacle for the reactor core fragments, with a dome shaped part to distribute the molten fragments and at least one outside layer of alumina bricks around the dome. The characteristic of this receptacle is that the outer layer of bricks contains neutron poison rods which pass through the bricks and protrude in relation to them [fr

  3. Organizing Core Tasks

    DEFF Research Database (Denmark)

    Boll, Karen

    has remained much the same within the last 10 years. However, how the core task has been organized has changed considerable under the influence of various “organizing devices”. The paper focusses on how organizing devices such as risk assessment, output-focus, effect orientation, and treatment...... projects influence the organization of core tasks within the tax administration. The paper shows that the organizational transformations based on the use of these devices have had consequences both for the overall collection of revenue and for the employees’ feeling of “making a difference”. All in all...

  4. GREEN CORE HOUSE

    Directory of Open Access Journals (Sweden)

    NECULAI Oana

    2017-05-01

    Full Text Available The Green Core House is a construction concept with low environmental impact, having as main central element a greenhouse. The greenhouse has the innovative role to use the biomass energy provided by plants to save energy. Although it is the central piece, the greenhouse is not the most innovative part of the Green Core House, but the whole building ensemble because it integrates many other sustainable systems as "waste purification systems", "transparent photovoltaic panels" or "double skin façades".

  5. PWR core design calculations

    Energy Technology Data Exchange (ETDEWEB)

    Trkov, A; Ravnik, M; Zeleznik, N [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    Functional description of the programme package Cord-2 for PWR core design calculations is presented. Programme package is briefly described. Use of the package and calculational procedures for typical core design problems are treated. Comparison of main results with experimental values is presented as part of the verification process. (author) [Slovenian] Opisali smo programski paket CORD-2, ki se uporablja pri projektnih izracunih sredice pri upravljanju tlacnovodnega reaktorja. Prikazana je uporaba paketa in racunskih postopkov za tipicne probleme, ki nastopajo pri projektiranju sredice. Primerjava glavnih rezultatov z eksperimentalnimi vrednostmi je predstavljena kot del preveritvenega procesa. (author)

  6. Ordered bulk degradation via autophagy

    DEFF Research Database (Denmark)

    Dengjel, Jörn; Kristensen, Anders Riis; Andersen, Jens S

    2008-01-01

    During amino acid starvation, cells undergo macroautophagy which is regarded as an unspecific bulk degradation process. Lately, more and more organelle-specific autophagy subtypes such as reticulophagy, mitophagy and ribophagy have been described and it could be shown, depending on the experimental...... at proteasomal and lysosomal degradation ample cross-talk between the two degradation pathways became evident. Degradation via autophagy appeared to be ordered and regulated at the protein complex/organelle level. This raises several important questions such as: can macroautophagy itself be specific and what...

  7. Degradation of thiram in soil

    International Nuclear Information System (INIS)

    Raghu, K.; Murthy, N.B.K.; Kumarsamy, R.

    1975-01-01

    Determination of the residual 35 S labelled tetramethylthiuram disulfide showed that the fungicide persisted longer in sterilized than in unsterilized soil, while the chloroform extractable radioactivity decreased, the water extractable radioactivity increased with increase in time. However, in sterilized soil the water extractable radioactivity remained more or less constant. Degradation of the fungicide was further demonstrated by the release of C 35 S 2 from soil treated with labelled thiram. Dimethylamine was found to be one of the degradation products. A bacterium isolated from thiram-enriched soil could degrade the fungicide in shake culture. The degradation pathways of thiram in sterilized and unsterilized soils are discussed. (author)

  8. In vitro degradation of ribosomes.

    Science.gov (United States)

    Mora, G; Rivas, A

    1976-12-01

    The cytoplasmic ribosomes from Euglena gracilis var. bacillaris are found to be of two types taking into consideration their stability "in vitro". In the group of unstable ribosomes the large subunit is degraded. The other group apparently does not suffer any degradation under the conditions described. However the RNAs extracted from both types of ribosomes are degraded during sucrose density gradients. The degradation of the largest RNA species has been reported previously, but no comment has been made about the stability of the ribosome itself.

  9. The radiation degradation of polypropylene

    International Nuclear Information System (INIS)

    De Hollain, G.

    1977-04-01

    Polypropylene is used extensively in the manufacture of disposable medical devices because of its superior properties. Unfortunately this polymer does not lend itself well to radiation sterilization, undergoing serious degradation which affects the mechanical properties of the polymer. In this paper the effects of radiation on the mechanical and physical properties of polypropylene are discussed. A programme of research to minimize the radiation degradation of this polymer through the addition of crosslinking agents to counteract the radiation degradation is proposed. It is furthermore proposed that a process of annealing of the irradiated polymer be investigated in order to minimize the post-irradiation degradation of the polypropylene [af

  10. Self-Healing Many-Core Architecture: Analysis and Evaluation

    Directory of Open Access Journals (Sweden)

    Arezoo Kamran

    2016-01-01

    Full Text Available More pronounced aging effects, more frequent early-life failures, and incomplete testing and verification processes due to time-to-market pressure in new fabrication technologies impose reliability challenges on forthcoming systems. A promising solution to these reliability challenges is self-test and self-reconfiguration with no or limited external control. In this work a scalable self-test mechanism for periodic online testing of many-core processor has been proposed. This test mechanism facilitates autonomous detection and omission of faulty cores and makes graceful degradation of the many-core architecture possible. Several test components are incorporated in the many-core architecture that distribute test stimuli, suspend normal operation of individual processing cores, apply test, and detect faulty cores. Test is performed concurrently with the system normal operation without any noticeable downtime at the application level. Experimental results show that the proposed test architecture is extensively scalable in terms of hardware overhead and performance overhead that makes it applicable to many-cores with more than a thousand processing cores.

  11. Maximum stellar iron core mass

    Indian Academy of Sciences (India)

    An analytical method of estimating the mass of a stellar iron core, just prior to core collapse, is described in this paper. The method employed depends, in part, upon an estimate of the true relativistic mass increase experienced by electrons within a highly compressed iron core, just prior to core collapse, and is significantly ...

  12. Nuclear core baffling apparatus

    International Nuclear Information System (INIS)

    Cooper, F.W. Jr.; Silverblatt, B.L.; Knight, C.B.; Berringer, R.T.

    1979-01-01

    An apparatus for baffling the flow of reactor coolant fluid into and about the core of a nuclear reactor is described. The apparatus includes a plurality of longitudinally aligned baffle plates with mating surfaces that allow longitudinal growth with temperature increases while alleviating both leakage through the aligned plates and stresses on the components supporting the plates

  13. The Uncommon Core

    Science.gov (United States)

    Ohler, Jason

    2013-01-01

    This author contends that the United States neglects creativity in its education system. To see this, he states, one may look at the Common Core State Standards. If one searches the English Language Arts and Literacy standards for the words "creative," "innovative," and "original"--and any associated terms, one will…

  14. Utah's New Mathematics Core

    Science.gov (United States)

    Utah State Office of Education, 2011

    2011-01-01

    Utah has adopted more rigorous mathematics standards known as the Utah Mathematics Core Standards. They are the foundation of the mathematics curriculum for the State of Utah. The standards include the skills and understanding students need to succeed in college and careers. They include rigorous content and application of knowledge and reflect…

  15. Some Core Contested Concepts

    Science.gov (United States)

    Chomsky, Noam

    2015-01-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and…

  16. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

    Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

  17. Investigation of EAS cores

    Directory of Open Access Journals (Sweden)

    Shaulov S.B.

    2017-01-01

    Full Text Available The development of nuclear-electromagnetic cascade models in air in the late forties have shown informational content of the study of cores of extensive air showers (EAS. These investigations were the main goal in different experiments which were carried out over many years by a variety of methods. Outcomes of such investigations obtained in the HADRON experiment using an X-ray emulsion chamber (XREC as a core detector are considered. The Ne spectrum of EAS associated with γ-ray families, spectra of γ-rays (hadrons in EAS cores and the Ne dependence of the muon number, ⟨Nμ⟩, in EAS with γ-ray families are obtained for the first time at energies of 1015–1017 eV with this method. A number of new effects were observed, namely, an abnormal scaling violation in hadron spectra which are fundamentally different from model predictions, an excess of muon number in EAS associated with γ-ray families, and the penetrating component in EAS cores. It is supposed that the abnormal behavior of γ-ray spectra and Ne dependence of the muon number are explained by the emergence of a penetrating component in the 1st PCR spectrum ‘knee’ range. Nuclear and astrophysical explanations of the origin of the penetrating component are discussed. The necessity of considering the contribution of a single close cosmic-ray source to explain the PCR spectrum in the knee range is noted.

  18. Plutonium cores of zenith

    Energy Technology Data Exchange (ETDEWEB)

    Barclay, F R; Cameron, I R; Drageset, A; Freemantle, R G; Wilson, D J

    1965-03-15

    The report describes a series of experiments carried out with plutonium fuel in the heated zero power reactor ZENITH, with the aim of testing current theoretical methods, with particular reference to excess reactivity, temperature coefficients, differential spectrum and reaction rate distributions. Two cores of widely different fissile/moderator atom ratios were loaded in order to test the theory under significantly varied spectrum conditions.

  19. Core damage risk indicators

    International Nuclear Information System (INIS)

    Szikszai, T.

    1994-01-01

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

  20. Core calculations of JMTR

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  1. Emergency core cooling system

    International Nuclear Information System (INIS)

    Kato, Ken.

    1989-01-01

    In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

  2. Core loss during a severe accident (COLOSS)

    International Nuclear Information System (INIS)

    Adroguer, B.; Bertrand, F.; Chatelard, P.; Cocuaud, N.; Van Dorsselaere, J.P.; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtilkova, V.; Belovsky, L.; Mueller, K.; Hering, W.; Homann, C.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrueck, M.; Stuckert, J.; Hozer, Z.; Bandini, G.; Birchley, J.; Berlepsch, T. von; Kleinhietpass, I.; Buck, M.; Benitez, J.A.F.; Virtanen, E.; Marguet, S.; Azarian, G.; Caillaux, A.; Plank, H.; Boldyrev, A.; Veshchunov, M.; Kobzar, V.; Zvonarev, Y.; Goryachev, A.

    2005-01-01

    The COLOSS project was a 3-year shared-cost action, which started in February 2000. The work-programme performed by 19 partners was shaped around complementary activities aimed at improving severe accident codes. Unresolved risk-relevant issues regarding H 2 production, melt generation and the source term were studied through a large number of experiments such as (a) dissolution of fresh and high burn-up UO 2 and MOX by molten Zircaloy (b) simultaneous dissolution of UO 2 and ZrO 2 (c) oxidation of U-O-Zr mixtures (d) degradation-oxidation of B 4 C control rods. Corresponding models were developed and implemented in severe accident computer codes. Upgraded codes were then used to apply results in plant calculations and evaluate their consequences on key severe accident sequences in different plants involving B 4 C control rods and in the TMI-2 accident. Significant results have been produced from separate-effects, semi-global and large-scale tests on COLOSS topics enabling the development and validation of models and the improvement of some severe accident codes. Breakthroughs were achieved on some issues for which more data are needed for consolidation of the modelling in particular on burn-up effects on UO 2 and MOX dissolution and oxidation of U-O-Zr and B 4 C-metal mixtures. There was experimental evidence that the oxidation of these mixtures can contribute significantly to the large H 2 production observed during the reflooding of degraded cores under severe accident conditions. The plant calculation activity enabled (a) the assessment of codes to calculate core degradation with the identification of main uncertainties and needs for short-term developments and (b) the identification of safety implications of new results. Main results and recommendations for future R and D activities are summarized in this paper

  3. Inflation targeting and core inflation

    OpenAIRE

    Julie Smith

    2005-01-01

    This paper examines the interaction of core inflation and inflation targeting as a monetary policy regime. Interest in core inflation has grown because of inflation targeting. Core inflation is defined in numerous ways giving rise to many potential measures; this paper defines core inflation as the best forecaster of inflation. A cross-country study finds before the start of inflation targeting, but not after, core inflation differs between non-inflation targeters and inflation targeters. Thr...

  4. CORE annual report 2006; CORE Jahresbericht 2006

    Energy Technology Data Exchange (ETDEWEB)

    Gut, A

    2007-04-15

    This annual report for the Swiss Federal Office of Energy (SFOE) summarises the activities of the Swiss Federal Commission on Energy Research CORE in 2006. The six main areas of work during the period 2004 - 2007 are examined, including a review of the SFOE's energy research programme, a road-map for the way towards the realisation of a 2000-watt society, the formulation of an energy research concept for 2008 - 2011, international co-operation, the dissemination of information and the assessment of existing and new instruments. International activities and Switzerland's involvement in energy research within the framework of the International Energy Agency IEA are discussed. New and existing projects are listed and the work done at the Competence Centre for Energy and Mobility noted. The Swiss Technology Award 2007 is presented. Information supplied to interested bodies to help improve knowledge on research work being done and to help make discussions on future energy supply more objective is discussed.

  5. DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

    Energy Technology Data Exchange (ETDEWEB)

    J. CUNNANE

    2004-11-19

    Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an &apos

  6. Polycarbonate radiolytic degradation and stabilization

    International Nuclear Information System (INIS)

    Araujo, E.S. de

    1994-01-01

    Polycarbonate Durolon, useful for medical supplies fabrication, is submitted to gamma radiation for sterilization purposes. Scissions in main chain occur, in carbonyl groups, producing molecular degradations and yellowness. The radiolytic stabilization is obtained through additive to the polymer. In this work some degradation and stabilization aspects are presented. (L.C.J.A.). 7 refs, 7 figs, 2 tabs

  7. Degradation of copepod fecal pellets

    DEFF Research Database (Denmark)

    Poulsen, Louise K.; Iversen, Morten

    2008-01-01

    amount of fecal pellets. The total degradation rate of pellets by the natural plankton community of Oresund followed the phytoplankton biomass, with maximum degradation rate during the spring bloom (2.5 +/- 0.49 d(-1)) and minimum (0.52 +/- 0.14 d(-1)) during late winter. Total pellet removal rate ranged...

  8. Degradable polymers for tissue engineering

    NARCIS (Netherlands)

    van Dijkhuizen-Radersma, Riemke; Moroni, Lorenzo; van Apeldoorn, Aart A.; Zhang, Zheng; Grijpma, Dirk W.; van Blitterswijk, Clemens A.

    2008-01-01

    This chapter elaborates the degradable polymers for tissue engineering and their required scaffold material in tissue engineering. It recognizes the examples of degradable polymers broadly used in tissue engineering. Tissue engineering is the persuasion of the body to heal itself through the

  9. MOSFET Degradation Under RF Stress

    NARCIS (Netherlands)

    Sasse, G.T.; Kuper, F.G.; Schmitz, Jurriaan

    2008-01-01

    We report on the degradation of MOS transistors under RF stress. Hot-carrier degradation, negative-bias temperature instability, and gate dielectric breakdown are investigated. The findings are compared to established voltage- and field-driven models. The experimental results indicate that the

  10. Designs for degraded Trbovlje

    Directory of Open Access Journals (Sweden)

    Naja Marot

    2005-01-01

    Full Text Available As an introduction, two degraded urban areas are presented. The first, planning unit seven, is situated in the southeastern part of Trbovlje town. The other, called Speke, lies to the south of Liverpool. The basis for the concept and context of urban renewal model are given by comparison between the newest Slovene and British spatial planning legislation, analyses of the Design management plan Nasipi and Supplementary Planning Document Edge Lane West, and review of different approaches to local communities’ involvement. Based on all the thus far collected data, a questionnaire about quality of living, knowledge of planning system and area perception was produced. Initially, it was used in a pilot residential area Žabjek, and afterwards, a shortened version was carried out in units lying in other parts of the town. Other stakeholders also expressed their ideas about how to develop planning unit seven. Speke Garston as another example of successful urban renewal is given. In conclusion guidelines for method and context development of urban renewal are given for planning unit seven, with emphasis on the Žabjek estate.

  11. Lysosomal degradation of membrane lipids.

    Science.gov (United States)

    Kolter, Thomas; Sandhoff, Konrad

    2010-05-03

    The constitutive degradation of membrane components takes place in the acidic compartments of a cell, the endosomes and lysosomes. Sites of lipid degradation are intralysosomal membranes that are formed in endosomes, where the lipid composition is adjusted for degradation. Cholesterol is sorted out of the inner membranes, their content in bis(monoacylglycero)phosphate increases, and, most likely, sphingomyelin is degraded to ceramide. Together with endosomal and lysosomal lipid-binding proteins, the Niemann-Pick disease, type C2-protein, the GM2-activator, and the saposins sap-A, -B, -C, and -D, a suitable membrane lipid composition is required for degradation of complex lipids by hydrolytic enzymes. Copyright 2009 Federation of European Biochemical Societies. Published by Elsevier B.V. All rights reserved.

  12. Ice cores and palaeoclimate

    International Nuclear Information System (INIS)

    Krogh Andersen, K.; Ditlevsen, P.; Steffensen, J.P.

    2001-01-01

    Ice cores from Greenland give testimony of a highly variable climate during the last glacial period. Dramatic climate warmings of 15 to 25 deg. C for the annual average temperature in less than a human lifetime have been documented. Several questions arise: Why is the Holocene so stable? Is climatic instability only a property of glacial periods? What is the mechanism behind the sudden climate changes? Are the increased temperatures in the past century man-made? And what happens in the future? The ice core community tries to attack some of these problems. The NGRIP ice core currently being drilled is analysed in very high detail, allowing for a very precise dating of climate events. It will be possible to study some of the fast changes on a year by year basis and from this we expect to find clues to the sequence of events during rapid changes. New techniques are hoped to allow for detection of annual layers as far back as 100,000 years and thus a much improved time scale over past climate changes. It is also hoped to find ice from the Eemian period. If the Eemian layers confirm the GRIP sequence, the Eemian was actually climatically unstable just as the glacial period. This would mean that the stability of the Holocene is unique. It would also mean, that if human made global warming indeed occurs, we could jeopardize the Holocene stability and create an unstable 'Eemian situation' which ultimately could start an ice age. Currenlty mankind is changing the composition of the atmosphere. Ice cores document significant increases in greenhouse gases, and due to increased emissions of sulfuric and nitric acid from fossil fuel burning, combustion engines and agriculture, modern Greenland snow is 3 - 5 times more acidic than pre-industrial snow (Mayewski et al., 1986). However, the magnitude and abruptness of the temperature changes of the past century do not exceed the magnitude of natural variability. It is from the ice core perspective thus not possible to attribute the

  13. Analysis of effect of cable degradation on SPND IR calculation

    International Nuclear Information System (INIS)

    Tamboli, P.K.; Sharma, A.; Prasad, A.D.; Singh, Nita; Antony, J.; Kelkar, M.G.; Kaurav, Reetesh; Pramanik, M.

    2013-01-01

    Neutron flux is the most vital parameter in the nuclear reactor safety against Neutronic over power. The modern days Indian PHWRs with large core size are loosely coupled reactors and hence In-core Self Power Neutron Detectors (SPNDs) are most suitable for monitoring local neutron power for generating Regional Overpower Trip. However the SPNDs and its Mineral Insulation Cable are prone to IR loss due to use of ceramic insulation which are highly hygroscopic. The present paper covers the online analysis of IR f degraded cable as per the surveillance requirement of monitoring the IR to assess the healthiness of SPNDs which are part of SSC/SSE for Reactor Protection Systems. The paper also proposes an alternative method for monitoring IR for startup//low power range when SPND signals are yet to pick up and Reactor Control and Protection are based on out of core Ionization Chambers. (author)

  14. Emergency core cooling device

    International Nuclear Information System (INIS)

    Suzaki, Kiyoshi; Inoue, Akihiro.

    1979-01-01

    Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

  15. Plasma core reactor applications

    International Nuclear Information System (INIS)

    Latham, T.S.; Rodgers, R.J.

    1976-01-01

    Analytical and experimental investigations are being conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride (UF 6 ) fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Power, in the form of thermal radiation emitted from the high-temperature nuclear fuel, is transmitted through fused-silica transparent walls to working fluids which flow in axial channels embedded in segments of the cavity walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration; each cavity is approximately 1 m in diameter by 4.35 m in length. Axial working fluid channels are located along a fraction of each cavity peripheral wall

  16. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

    Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

  17. Drift Degradation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Dwayne C. Kicker

    2001-09-28

    A statistical description of the probable block sizes formed by fractures around the emplacement drifts has been developed for each of the lithologic units of the repository host horizon. A range of drift orientations with the drift azimuth varied in 15{sup o} increments has been considered in the static analysis. For the quasi-static seismic analysis, and the time-dependent and thermal effects analysis, two drift orientations have been considered: a drift azimuth of 105{sup o} and the current emplacement drift azimuth of 75{sup o}. The change in drift profile resulting from progressive deterioration of the emplacement drifts has been assessed both with and without backfill. Drift profiles have been determined for four different time increments, including static (i.e., upon excavation), 200 years, 2,000 years, and 10,000 years. The effect of seismic events on rock fall has been analyzed. Block size distributions and drift profiles have been determined for three seismic levels, including a 1,000-year event, a 5,000-year event, and a 10,000-year event. Data developed in this modeling and analysis activity have been entered into the TDMS (DTN: MO0109RDDAAMRR.003). The following conclusions have resulted from this drift degradation analysis: (1) The available fracture data are suitable for supporting a detailed key block analysis of the repository host horizon rock mass. The available data from the north-south Main Drift and the east-west Cross Drift provide a sufficient representative fracture sample of the repository emplacement drift horizon. However, the Tptpln fracture data are only available from a relatively small section of the Cross Drift, resulting in a smaller fracture sample size compared to the other lithologic units. This results in a lower degree of confidence that the key block data based on the Tptpln data set is actually representative of the overall Tptpln key block population. (2) The seismic effect on the rock fall size distribution for all events

  18. Some core contested concepts.

    Science.gov (United States)

    Chomsky, Noam

    2015-02-01

    Core concepts of language are highly contested. In some cases this is legitimate: real empirical and conceptual issues arise. In other cases, it seems that controversies are based on misunderstanding. A number of crucial cases are reviewed, and an approach to language is outlined that appears to have strong conceptual and empirical motivation, and to lead to conclusions about a number of significant issues that differ from some conventional beliefs.

  19. Schumpeter's core works revisited

    DEFF Research Database (Denmark)

    Andersen, Esben Sloth

    2012-01-01

    This paper organises Schumpeter’s core books in three groups: the programmatic duology,the evolutionaryeconomic duology,and the socioeconomic synthesis. By analysing these groups and their interconnections from the viewpoint of modern evolutionaryeconomics,the paper summarises resolved problems a...... and points at remaining challenges. Its analyses are based on distinctions between microevolution and macroevolution, between economic evolution and socioeconomic coevolution, and between Schumpeter’s three major evolutionary models (called Mark I, Mark II and Mark III)....

  20. BWR type reactor core

    International Nuclear Information System (INIS)

    Tatemichi, Shin-ichiro.

    1981-01-01

    Purpose: To eliminate the variation in the power distribution of a BWR type reactor core in the axial direction even if the flow rate is increased or decreased by providing a difference in the void coefficient between the upper part and the lower parts of the reactor core, and increasing the void coefficient at the lower part of the reactor core. Constitution: The void coefficient of the lower region from the center to the lower part along the axial direction of a nuclear fuel assembly is increased to decrease the dependence on the flow rate of the axial power distribution of the nuclear fuel assembly. That is, a water/fuel ratio is varied, the water in non-boiled region is increased or the neutron spectrum is varied so as to vary the void coefficient. In order to exemplify it, the rate of the internal pellets of the fuel rod of the nuclear fuel assembly or the shape of the channel box is varied. Accordingly, the power does not considerably vary even if the flow rate is altered since the power is varied in the power operation. (Yoshihara, H.)

  1. Emergency core cooling system

    International Nuclear Information System (INIS)

    Ando, Masaki.

    1987-01-01

    Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

  2. Identification of chlorinated solvents degradation zones in clay till by high resolution chemical, microbial and compound specific isotope analysis

    DEFF Research Database (Denmark)

    Damgaard, Ida; Bjerg, Poul Løgstrup; Bælum, Jacob

    2013-01-01

    subsampling of the clay till cores. The study demonstrates that an integrated approach combining chemical analysis, molecular microbial tools and compound specific isotope analysis (CSIA) was required in order to document biotic and abiotic degradations in the clay till system. © 2013 Elsevier B.V.......The degradation of chlorinated ethenes and ethanes in clay till was investigated at a contaminated site (Vadsby, Denmark) by high resolution sampling of intact cores combined with groundwater sampling. Over decades of contamination, bioactive zones with degradation of trichloroethene (TCE) and 1...

  3. Development of proactive technology against nuclear materials degradation

    International Nuclear Information System (INIS)

    Jeong, Yong Hwan; Kim, Hong Pyo; Lee, Bong Sang

    2012-04-01

    As the nuclear power plants are getting older, the extent of materials degradation increases and unexpected degradation mechanisms may occur under complex environments, including high-temperature and pressure, radiation and coolant. The components in the primary system are maintained at the temperature of 320 .deg. C, pressure of 2500 psi, and reactor internals are exposed to fast neutrons. The pipes and nozzles are affected by the mechanical, thermal and corrosive cyclic fatigue stresses. Since the steam generator tubes are affected by both primary and secondary coolants, the materials degradation mechanisms are dependent upon the multiple or complex factors. In this report, we make contribution to the enhancement of reactor safety by developing techniques for predicting and evaluating materials behaviors in nuclear environments. The research product in the following five areas, described in this report, plays a vital role in improving the safe operation of nuclear reactors, upgrading the level of skills and extending the use of nuclear power. Development of corrosion control and protection technology Development of fracture mechanical evaluation model of reactor pressure Development of prediction and analysis technology for radiation damage Development of advanced diagnostic techniques for micro-materials degradation Development of core technology for control of steam generator degradation

  4. Intrinsic immunogenicity of rapidly-degradable polymers evolves during degradation.

    Science.gov (United States)

    Andorko, James I; Hess, Krystina L; Pineault, Kevin G; Jewell, Christopher M

    2016-03-01

    Recent studies reveal many biomaterial vaccine carriers are able to activate immunostimulatory pathways, even in the absence of other immune signals. How the changing properties of polymers during biodegradation impact this intrinsic immunogenicity is not well studied, yet this information could contribute to rational design of degradable vaccine carriers that help direct immune response. We use degradable poly(beta-amino esters) (PBAEs) to explore intrinsic immunogenicity as a function of the degree of polymer degradation and polymer form (e.g., soluble, particles). PBAE particles condensed by electrostatic interaction to mimic a common vaccine approach strongly activate dendritic cells, drive antigen presentation, and enhance T cell proliferation in the presence of antigen. Polymer molecular weight strongly influences these effects, with maximum stimulation at short degradation times--corresponding to high molecular weight--and waning levels as degradation continues. In contrast, free polymer is immunologically inert. In mice, PBAE particles increase the numbers and activation state of cells in lymph nodes. Mechanistic studies reveal that this evolving immunogenicity occurs as the physicochemical properties and concentration of particles change during polymer degradation. This work confirms the immunological profile of degradable, synthetic polymers can evolve over time and creates an opportunity to leverage this feature in new vaccines. Degradable polymers are increasingly important in vaccination, but how the inherent immunogenicity of polymers changes during degradation is poorly understood. Using common rapidly-degradable vaccine carriers, we show that the activation of immune cells--even in the absence of other adjuvants--depends on polymer form (e.g., free, particulate) and the extent of degradation. These changing characteristics alter the physicochemical properties (e.g., charge, size, molecular weight) of polymer particles, driving changes in

  5. Comparison of the fractional power motor with cores made of various magnetic materials

    Science.gov (United States)

    Gmyrek, Zbigniew; Lefik, Marcin; Cavagnino, Andrea; Ferraris, Luca

    2017-12-01

    The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite), reduce the core loss and/or provide quasi-isotropic core's properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

  6. Operationalizing measurement of forest degradation

    DEFF Research Database (Denmark)

    Dons, Klaus; Smith-Hall, Carsten; Meilby, Henrik

    2015-01-01

    . In Tanzania, charcoal production is considered a major cause of forest degradation, but is challenging to quantify due to sub-canopy biomass loss, remote production sites and illegal trade. We studied two charcoal production sites in dry Miombo woodland representing open woodland conditions near human......Quantification of forest degradation in monitoring and reporting as well as in historic baselines is among the most challenging tasks in national REDD+ strategies. However, a recently introduced option is to base monitoring systems on subnational conditions such as prevalent degradation activities...

  7. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  8. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

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

  9. WNP-2 core model upgrade

    International Nuclear Information System (INIS)

    Golightly, C.E.; Ravindranath, T.K.; Belblidia, L.A.; O'Farrell, D.; Andersen, P.S.

    2006-01-01

    The paper describes the core model upgrade of the WNP-2 training simulator and the reasons for the upgrade. The core model as well as the interface with the rest of the simulator are briefly described . The paper also describes the procedure that will be used by WNP-2 to update the simulator core data after future core reloads. Results from the fully integrated simulator are presented. (author)

  10. On-line core monitoring with CORE MASTER / PRESTO

    International Nuclear Information System (INIS)

    Lindahl, S.O.; Borresen, S.; Ovrum, S.

    1986-01-01

    Advanced calculational tools are instrumental in improving reactor plant capacity factors and fuel utilization. The computer code package CORE MASTER is an integrated system designed to achieve this objective. The system covers all main activities in the area of in-core fuel management for boiling water reactors; design, operation support, and on-line core monitoring. CORE MASTER operates on a common data base, which defines the reactor and documents the operating history of the core and of all fuel bundles ever used

  11. Ecosystemic approaches to land degradation

    Energy Technology Data Exchange (ETDEWEB)

    Puigdefabregas, J.; Barrio, G. del; Hill, J.

    2009-07-01

    Land degradation is recognized as the main outcome of desertification. However available procedures for its assessment are still unsatisfactory because are often too costly for surveying large areas and rely on specific components of the degradation process without being able to integrate them in a unique process. One of the objectives of De Survey project is designing and implementing operational procedures for desertification surveillance, including land degradation. A strategic report was compiled and reproduced here for selecting the most appropriate approaches to the project conditions. The report focuses on using attributes of ecosystem maturity as a natural way to integrate the different drivers of land degradation in simple indices. The review surveys different families of attributes concerned with water and energy fluxes through the ecosystem, its capacity to sustain biomass and net primary productivity, and its capacity to structure the space. Finally, some conclusions are presented about the choice criteria of the different approaches in the framne of operational applications. (Author) 20 refs.

  12. Ecosystemic approaches to land degradation

    International Nuclear Information System (INIS)

    Puigdefabregas, J.; Barrio, G. del; Hill, J.

    2009-01-01

    Land degradation is recognized as the main outcome of desertification. However available procedures for its assessment are still unsatisfactory because are often too costly for surveying large areas and rely on specific components of the degradation process without being able to integrate them in a unique process. One of the objectives of De Survey project is designing and implementing operational procedures for desertification surveillance, including land degradation. A strategic report was compiled and reproduced here for selecting the most appropriate approaches to the project conditions. The report focuses on using attributes of ecosystem maturity as a natural way to integrate the different drivers of land degradation in simple indices. The review surveys different families of attributes concerned with water and energy fluxes through the ecosystem, its capacity to sustain biomass and net primary productivity, and its capacity to structure the space. Finally, some conclusions are presented about the choice criteria of the different approaches in the framne of operational applications. (Author) 20 refs.

  13. Chitin Degradation In Marine Bacteria

    DEFF Research Database (Denmark)

    Paulsen, Sara; Machado, Henrique; Gram, Lone

    2015-01-01

    Introduction: Chitin is the most abundant polymer in the marine environment and the second most abundant in nature. Chitin does not accumulate on the ocean floor, because of microbial breakdown. Chitin degrading bacteria could have potential in the utilization of chitin as a renewable carbon...... and nitrogen source in the fermentation industry.Methods: Here, whole genome sequenced marine bacteria were screened for chitin degradation using phenotypic and in silico analyses.Results: The in silico analyses revealed the presence of three to nine chitinases in each strain, however the number of chitinases...... chitin regulatory system.Conclusions: This study has provided insight into the ecology of chitin degradation in marine bacteria. It also served as a basis for choosing a more efficient chitin degrading production strain e.g. for the use of chitin waste for large-scale fermentations....

  14. Predicting degradability of organic chemicals

    Energy Technology Data Exchange (ETDEWEB)

    Finizio, A; Vighi, M [Milan Univ. (Italy). Ist. di Entomologia Agraria

    1992-05-01

    Degradability, particularly biodegradability, is one of the most important factors governing the persistence of pollutants in the environment and consequently influencing their behavior and toxicity in aquatic and terrestrial ecosystems. The need for reliable persistence data in order to assess the environmental fate and hazard of chemicals by means of predictive approaches, is evident. Biodegradability tests are requested by the EEC directive on new chemicals. Neverthless, degradation tests are not easy to carry out and data on existing chemicals are very scarce. Therefore, assessing the fate of chemicals in the environment from the simple study of their structure would be a useful tool. Rates of degradation are a function of the rates of a series of processes. Correlation between degradation rates and structural parameters are will be facilitated if one of the processes is rate determining. This review is a survey of studies dealing with relationships between structure and biodegradation of organic chemicals, to identify the value and limitations of this approach.

  15. Hollow-Core Fiber Lamp

    Science.gov (United States)

    Yi, Lin (Inventor); Tjoelker, Robert L. (Inventor); Burt, Eric A. (Inventor); Huang, Shouhua (Inventor)

    2016-01-01

    Hollow-core capillary discharge lamps on the millimeter or sub-millimeter scale are provided. The hollow-core capillary discharge lamps achieve an increased light intensity ratio between 194 millimeters (useful) and 254 millimeters (useless) light than conventional lamps. The capillary discharge lamps may include a cone to increase light output. Hollow-core photonic crystal fiber (HCPCF) may also be used.

  16. Dual-core Itanium Processor

    CERN Multimedia

    2006-01-01

    Intel’s first dual-core Itanium processor, code-named "Montecito" is a major release of Intel's Itanium 2 Processor Family, which implements the Intel Itanium architecture on a dual-core processor with two cores per die (integrated circuit). Itanium 2 is much more powerful than its predecessor. It has lower power consumption and thermal dissipation.

  17. Maximum stellar iron core mass

    Indian Academy of Sciences (India)

    60, No. 3. — journal of. March 2003 physics pp. 415–422. Maximum stellar iron core mass. F W GIACOBBE. Chicago Research Center/American Air Liquide ... iron core compression due to the weight of non-ferrous matter overlying the iron cores within large .... thermal equilibrium velocities will tend to be non-relativistic.

  18. Core TuLiP

    NARCIS (Netherlands)

    Czenko, M.R.; Etalle, Sandro

    2007-01-01

    We propose CoreTuLiP - the core of a trust management language based on Logic Programming. CoreTuLiP is based on a subset of moded logic programming, but enjoys the features of TM languages such as RT; in particular clauses are issued by different authorities and stored in a distributed manner. We

  19. Automated Core Design

    International Nuclear Information System (INIS)

    Kobayashi, Yoko; Aiyoshi, Eitaro

    2005-01-01

    Multistate searching methods are a subfield of distributed artificial intelligence that aims to provide both principles for construction of complex systems involving multiple states and mechanisms for coordination of independent agents' actions. This paper proposes a multistate searching algorithm with reinforcement learning for the automatic core design of a boiling water reactor. The characteristics of this algorithm are that the coupling structure and the coupling operation suitable for the assigned problem are assumed and an optimal solution is obtained by mutual interference in multistate transitions using multiagents. Calculations in an actual plant confirmed that the proposed algorithm increased the convergence ability of the optimization process

  20. Ice Sheets & Ice Cores

    DEFF Research Database (Denmark)

    Mikkelsen, Troels Bøgeholm

    Since the discovery of the Ice Ages it has been evident that Earth’s climate is liable to undergo dramatic changes. The previous climatic period known as the Last Glacial saw large oscillations in the extent of ice sheets covering the Northern hemisphere. Understanding these oscillations known....... The first part concerns time series analysis of ice core data obtained from the Greenland Ice Sheet. We analyze parts of the time series where DO-events occur using the so-called transfer operator and compare the results with time series from a simple model capable of switching by either undergoing...

  1. Nuclear reactor core assembly

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1978-01-01

    The object of the present invention is to provide a fast reactor core assembly design for use with a fluid coolant such as liquid sodium or carbon monoxide incorporating a method of increasing the percentage of coolant flow though the blanket elements relative to the total coolant flow through the blanket and fuel elements during shutdown conditions without using moving parts. It is claimed that deterioration due to reactor radiation or temperature conditions is avoided and ready modification or replacement is possible. (U.K.)

  2. Reload core safety verification

    International Nuclear Information System (INIS)

    Svetlik, M.; Minarcin, M.

    2003-01-01

    This paper presents a brief look at the process of reload core safety evaluation and verification in Slovak Republic. It gives an overview of experimental verification of selected nuclear parameters in the course of physics testing during reactor start-up. The comparison of IAEA recommendations and testing procedures at Slovak and European nuclear power plants of similar design is included. An introduction of two level criteria for evaluation of tests represents an effort to formulate the relation between safety evaluation and measured values (Authors)

  3. RB reactor benchmark cores

    International Nuclear Information System (INIS)

    Pesic, M.

    1998-01-01

    A selected set of the RB reactor benchmark cores is presented in this paper. The first results of validation of the well-known Monte Carlo MCNP TM code and adjoining neutron cross section libraries are given. They confirm the idea for the proposal of the new U-D 2 O criticality benchmark system and support the intention to include this system in the next edition of the recent OECD/NEA Project: International Handbook of Evaluated Criticality Safety Experiment, in near future. (author)

  4. Working session 1: Tubing degradation

    International Nuclear Information System (INIS)

    Kharshafdjian, G.; Turluer, G.

    1997-01-01

    A general introductory overview of the purpose of the group and the general subject area of SG tubing degradation was given by the facilitator. The purpose of the session was described as to open-quotes develop conclusions and proposals on regulatory and technical needs required to deal with the issues of SG tubing degradation.close quotes Types, locations and characteristics of tubing degradation in steam generators were briefly reviewed. The well-known synergistic effects of materials, environment, and stress and strain/strain rate, subsequently referred to by the acronym open-quotes MESSclose quotes by some of the group members, were noted. The element of time (i.e., evolution of these variables with time) was emphasized. It was also suggested that the group might want to consider the related topics of inspection capabilities, operational variables, degradation remedies, and validity of test data, and some background information in these areas was provided. The presentation given by Peter Millet during the Plenary Session was reviewed; Specifically, the chemical aspects and the degradation from the secondary side of the steam generator were noted. The main issues discussed during the October 1995 EPRI meeting on secondary side corrosion were reported, and a listing of the potential SG tube degradations was provided and discussed

  5. Abiotic degradation of plastic films

    Science.gov (United States)

    Ángeles-López, Y. G.; Gutiérrez-Mayen, A. M.; Velasco-Pérez, M.; Beltrán-Villavicencio, M.; Vázquez-Morillas, A.; Cano-Blanco, M.

    2017-01-01

    Degradable plastics have been promoted as an option to mitigate the environmental impacts of plastic waste. However, there is no certainty about its degradability under different environmental conditions. The effect of accelerated weathering (AW), natural weathering (NW) and thermal oxidation (TO) on different plastics (high density polyethylene, HDPE; oxodegradable high density polyethylene, HDPE-oxo; compostable plastic, Ecovio ® metalized polypropylene, PP; and oxodegradable metalized polypropylene, PP-oxo) was studied. Plastics films were exposed to AW per 110 hours; to NW per 90 days; and to TO per 30 days. Plastic films exposed to AW and NW showed a general loss on mechanical properties. The highest reduction in elongation at break on AW occurred to HDPE-oxo (from 400.4% to 20.9%) and was higher than 90% for HDPE, HDPE-oxo, Ecovio ® and PP-oxo in NW. No substantial evidence of degradation was found on plastics exposed to TO. Oxo-plastics showed higher degradation rates than their conventional counterparts, and the compostable plastic was resistant to degradation in the studied abiotic conditions. This study shows that degradation of plastics in real life conditions will vary depending in both, their composition and the environment.

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

    Science.gov (United States)

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

    2018-03-01

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

  7. Monitoring an electric cable core

    International Nuclear Information System (INIS)

    Bhattacharya, S.; Marris, A.

    1984-01-01

    A method of, and apparatus for, continuously monitoring an advancing core having a continuous covering comprises directing X-ray radiation laterally towards the advancing covered core; continuously forming an X-ray image pattern of the advancing covered core and translating the image pattern into a visible image pattern; continuously transforming the visible pattern into a digital bit pattern; and processing the digital bit pattern using a microprocessor with interfacing electronics to provide an image profile of the advancing covered core and/or to provide analogue and/or digital signals indicative of the overall diameter and eccentricity of the covered core and of the thickness of the covering. (author)

  8. Winning Cores in Parity Games

    DEFF Research Database (Denmark)

    Vester, Steen

    2016-01-01

    We introduce the novel notion of winning cores in parity games and develop a deterministic polynomial-time under-approximation algorithm for solving parity games based on winning core approximation. Underlying this algorithm are a number properties about winning cores which are interesting...... in their own right. In particular, we show that the winning core and the winning region for a player in a parity game are equivalently empty. Moreover, the winning core contains all fatal attractors but is not necessarily a dominion itself. Experimental results are very positive both with respect to quality...

  9. Initial charge reactor core

    International Nuclear Information System (INIS)

    Kiyono, Takeshi

    1984-01-01

    Purpose: To effectivity burn fuels and improve the economical performance in an inital charge reactor core of BWR type reactors or the likes. Constitution: In a reactor core constituted with a plurality of fuel assemblies which are to be partially replaced upon fuel replacement, the density of the fissionable materials and the moderator - fuel ratio of a fuel assembly is set corresponding to the period till that fuel assembly is replaced, in which the density of the nuclear fissionable materials is lowered and the moderator - fuel ratio is increased for the fuel assembly with a shorter period from the fueling to the fuel exchange and, while on the other hand, the density of the fissionable materials is increased and the moderator - fuel ratio is decreased for the fuel assembly with a longer period from the fueling to the replacement. Accordingly, since the moderator - fuel ratio is increased for the fuel assembly to be replaced in a shorter period, the neutrons moderating effect is increased to increase the reactivity. (Horiuchi, T.)

  10. Statistical core design

    International Nuclear Information System (INIS)

    Oelkers, E.; Heller, A.S.; Farnsworth, D.A.; Kearfott, K.J.

    1978-01-01

    The report describes the statistical analysis of DNBR thermal-hydraulic margin of a 3800 MWt, 205-FA core under design overpower conditions. The analysis used LYNX-generated data at predetermined values of the input variables whose uncertainties were to be statistically combined. LYNX data were used to construct an efficient response surface model in the region of interest; the statistical analysis was accomplished through the evaluation of core reliability; utilizing propagation of the uncertainty distributions of the inputs. The response surface model was implemented in both the analytical error propagation and Monte Carlo Techniques. The basic structural units relating to the acceptance criteria are fuel pins. Therefore, the statistical population of pins with minimum DNBR values smaller than specified values is determined. The specified values are designated relative to the most probable and maximum design DNBR values on the power limiting pin used in present design analysis, so that gains over the present design criteria could be assessed for specified probabilistic acceptance criteria. The results are equivalent to gains ranging from 1.2 to 4.8 percent of rated power dependent on the acceptance criterion. The corresponding acceptance criteria range from 95 percent confidence that no pin will be in DNB to 99.9 percent of the pins, which are expected to avoid DNB

  11. Nuclear reactor core

    International Nuclear Information System (INIS)

    Koyama, Jun-ichi; Aoyama, Motoo; Ishibashi, Yoko; Mochida, Takaaki; Haikawa, Katsumasa; Yamanaka, Akihiro.

    1995-01-01

    A reactor core is radially divided into an inner region, an outer region and an outermost region. As a fuel, three kinds of fuels, namely, a high enrichment degree fuel at 3.4%, a middle enrichment degree fuel at 2.3% and a low enrichment degree at 1.1% of a fuel average enrichment degree of fission product are used. Each of the fuels is bisected to upper and lower portions at an axial center thereof. The difference of average enrichment degrees between upper and lower portions is 0.1% for the high enrichment degree fuel, 0.3% for the middle enrichment degree fuel and 0.2% for the low enrichment degree fuel. In addition, the composition of fuels in each of radial regions comprises 100% of the low enrichment degree fuels in the outermost region, 91% of the higher enrichment degree fuels and 9% of the middle enrichment degree fuels in the outer region, and 34% of the high enrichment degree fuels and 30% of the middle enrichment degree fuels in the inner region. With such a constitution, fuel economy can be improved while maintaining the thermal margin in an initially loaded reactor core of a BWR type reactor. (I.N.)

  12. Models of the earth's core

    Science.gov (United States)

    Stevenson, D. J.

    1981-01-01

    Combined inferences from seismology, high-pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to models of the earth's core with five basic properties. These are that core formation was contemporaneous with earth accretion; the core is not in chemical equilibrium with the mantle; the outer core is a fluid iron alloy containing significant quantities of lighter elements and is probably almost adiabatic and compositionally uniform; the more iron-rich inner solid core is a consequence of partial freezing of the outer core, and the energy release from this process sustains the earth's magnetic field; and the thermodynamic properties of the core are well constrained by the application of liquid-state theory to seismic and labroatory data.

  13. Overview of fuel behaviour and core degradation, based on modelling analyses. Overview of fuel behaviour and core degradation, on the basis of modelling results

    International Nuclear Information System (INIS)

    Massara, Simone

    2013-01-01

    Since the very first hours after the accident at Fukushima-Daiichi, numerical simulations by means of severe accident codes have been carried out, aiming at highlighting the key physical phenomena allowing a correct understanding of the sequence of events, and - on a long enough timeline - improving models and methods, in order to reduce the discrepancy between calculated and measured data. A last long-term objective is to support the future decommissioning phase. The presentation summarises some of the available elements on the role of the fuel/cladding-water interaction, which became available only through modelling because of the absence of measured data directly related to the cladding-steam interaction. This presentation also aims at drawing some conclusions on the status of the modelling capabilities of current tools, particularly for the purpose of the foreseen application to ATF fuels: - analyses with MELCOR, MAAP, THALES2 and RELAP5 are presented; - input data are taken from BWR Mark-I Fukushima-Daiichi Units 1, 2 and 3, completed with operational data published by TEPCO. In the case of missing or incomplete data or hypotheses, these are adjusted to reduce the calculation/measurement discrepancy. The behaviour of the accident is well understood on a qualitative level (major trends on RPV pressure and water level, dry-wet and PCV pressure are well represented), allowing a certain level of confidence in the results of the analysis of the zirconium-steam reaction - which is accessible only through numerical simulations. These show an extremely fast sequence of events (here for Unit 1): - the top of fuel is uncovered in 3 hours (after the tsunami); - the steam line breaks at 6.5 hours. Vessel dries at 10 hours, with a heat-up rate in a first moment driven by the decay heat only (∼7 K/min) and afterwards by the chemical heat from Zr-oxidation (over 30 K/min), associated with massive hydrogen production. It appears that the level of uncertainty increases with the progression of the accident (which is not surprising); if a good agreement among most of the numerical predictions and the plant parameter is observed in the pre-fuel melting phase, much higher uncertainties affect the post-melting phase and, ultimately, the final status of the RPV. Finally, nearly no elements allow confirming whether the capabilities of current severe accident tools could easily be extrapolated to accident-tolerant fuels. The major efforts should be focused on: - input data; - adaptation and improvement of physical models; - overall validation against experimental results; - The development of advanced modelling techniques, including multi-scale modelling

  14. Materials Degradation in Light Water Reactors: Life After 60,

    International Nuclear Information System (INIS)

    Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili; Naus, Dan J

    2008-01-01

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

  15. Waves in the core and mechanical core-mantle interactions

    DEFF Research Database (Denmark)

    Jault, D.; Finlay, Chris

    2015-01-01

    This Chapter focuses on time-dependent uid motions in the core interior, which can beconstrained by observations of the Earth's magnetic eld, on timescales which are shortcompared to the magnetic diusion time. This dynamics is strongly inuenced by the Earth's rapid rotation, which rigidies...... the motions in the direction parallel to the Earth'srotation axis. This property accounts for the signicance of the core-mantle topography.In addition, the stiening of the uid in the direction parallel to the rotation axis gives riseto a magnetic diusion layer attached to the core-mantle boundary, which would...... otherwisebe dispersed by Alfven waves. This Chapter complements the descriptions of large-scaleow in the core (8.04), of turbulence in the core (8.06) and of core-mantle interactions(8.12), which can all be found in this volume. We rely on basic magnetohydrodynamictheory, including the derivation...

  16. Core cooling systems

    International Nuclear Information System (INIS)

    Hoeppner, G.

    1980-01-01

    The reactor cooling system transports the heat liberated in the reactor core to the component - heat exchanger, steam generator or turbine - where the energy is removed. This basic task can be performed with a variety of coolants circulating in appropriately designed cooling systems. The choice of any one system is governed by principles of economics and natural policies, the design is determined by the laws of nuclear physics, thermal-hydraulics and by the requirement of reliability and public safety. PWR- and BWR- reactors today generate the bulk of nuclear energy. Their primary cooling systems are discussed under the following aspects: 1. General design, nuclear physics constraints, energy transfer, hydraulics, thermodynamics. 2. Design and performance under conditions of steady state and mild transients; control systems. 3. Design and performance under conditions of severe transients and loss of coolant accidents; safety systems. (orig./RW)

  17. The true 'core' splitting

    International Nuclear Information System (INIS)

    Hallerbach, J.

    1978-01-01

    Massive unemployment and the fear of a barred future put at present the unions and civil initiative to the apparent alternatives; securing work places or securing life and future. How the 'atomic fight' is fought and its result can have considerable consequences for our society. This volume presents a dialogue: Firstly the situation and environment must be understood giving rise to the controversial arguments. Reports, analyses and interviews are presented on this as basic structure for the future discussion. The quality and direction of the technical progress are dealt with in the core of the discussion. Is atomic technology acceptable. Who should decide and whom does it serve. What is progress going to look like anyway. (orig.) [de

  18. Emergency core cooling systems

    International Nuclear Information System (INIS)

    Kubokoya, Takashi; Okataku, Yasukuni.

    1984-01-01

    Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

  19. The core and cosmopolitans

    DEFF Research Database (Denmark)

    Dahlander, Linus; Frederiksen, Lars

    2012-01-01

    Users often interact and help each other solve problems in communities, but few scholars have explored how these relationships provide opportunities to innovate. We analyze the extent to which people positioned within the core of a community as well as people that are cosmopolitans positioned...... across multiple external communities affect innovation. Using a multimethod approach, including a survey, a complete database of interactions in an online community, content coding of interactions and contributions, and 36 interviews, we specify the types of positions that have the strongest effect...... on innovation. Our study shows that dispositional explanations for user innovation should be complemented by a relational view that emphasizes how these communities differ from other organizations, the types of behaviors this enables, and the effects on innovation....

  20. Adult educators' core competences

    DEFF Research Database (Denmark)

    Wahlgren, Bjarne

    2016-01-01

    ” requirements, organising them into four thematic subcategories: (1) communicating subject knowledge; (2) taking students’ prior learning into account; (3) supporting a learning environment; and (4) the adult educator’s reflection on his or her own performance. At the end of his analysis of different competence......Abstract Which competences do professional adult educators need? This research note discusses the topic from a comparative perspective, finding that adult educators’ required competences are wide-ranging, heterogeneous and complex. They are subject to context in terms of national and cultural...... environment as well as the kind of adult education concerned (e.g. basic education, work-related education etc.). However, it seems that it is possible to identify certain competence requirements which transcend national, cultural and functional boundaries. This research note summarises these common or “core...

  1. CORE annual report 2006

    International Nuclear Information System (INIS)

    Gut, A.

    2007-04-01

    This annual report for the Swiss Federal Office of Energy (SFOE) summarises the activities of the Swiss Federal Commission on Energy Research CORE in 2006. The six main areas of work during the period 2004 - 2007 are examined, including a review of the SFOE's energy research programme, a road-map for the way towards the realisation of a 2000-watt society, the formulation of an energy research concept for 2008 - 2011, international co-operation, the dissemination of information and the assessment of existing and new instruments. International activities and Switzerland's involvement in energy research within the framework of the International Energy Agency IEA are discussed. New and existing projects are listed and the work done at the Competence Centre for Energy and Mobility noted. The Swiss Technology Award 2007 is presented. Information supplied to interested bodies to help improve knowledge on research work being done and to help make discussions on future energy supply more objective is discussed

  2. Nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F

    1974-07-11

    The core of the fast neutron reactor consisting, among other components, of fuel elements enriched in plutonium is divided into modules. Each module contains a bundle of four or six elongated components (fuel elements and control rods). In the arrangement with four components, one is kept rigid while the other three are elastically yielding inclined towards the center and lean against the rigid component. In the modules with six pieces, each component is elastically yielding inclined towards a central cavity. In this way, they form a circular arc. A control rod may be placed in the cavity. In order to counteract a relative lateral movement, the outer surfaces of the components which have hexagonal cross-sections have interlocking bearing cushions. The bearing cushions consist of keyway-type ribs or grooves with the wedges or ribs gripping in the grooves of the neighbouring components. In addition, the ribs have oblique entering surfaces.

  3. IRIS core criticality calculations

    International Nuclear Information System (INIS)

    Jecmenica, R.; Trontl, K.; Pevec, D.; Grgic, D.

    2003-01-01

    Three-dimensional Monte Carlo computer code KENO-VI of CSAS26 sequence of SCALE-4.4 code system was applied for pin-by-pin calculations of the effective multiplication factor for the first cycle IRIS reactor core. The effective multiplication factors obtained by the above mentioned Monte Carlo calculations using 27-group ENDF/B-IV library and 238-group ENDF/B-V library have been compared with the effective multiplication factors achieved by HELIOS/NESTLE, CASMO/SIMULATE, and modified CORD-2 nodal calculations. The results of Monte Carlo calculations are found to be in good agreement with the results obtained by the nodal codes. The discrepancies in effective multiplication factor are typically within 1%. (author)

  4. Understanding core conductor fabrics

    International Nuclear Information System (INIS)

    Swenson, D E

    2011-01-01

    ESD Association standard test method ANSI/ESD STM2.1 - Garments (STM2.1), provides electrical resistance test procedures that are applicable for materials and garments that have surface conductive or surface dissipative properties. As has been reported in other papers over the past several years 1 fabrics are now used in many industries for electrostatic control purposes that do not have surface conductive properties and therefore cannot be evaluated using the procedures in STM2.1 2 . A study was conducted to compare surface conductive fabrics with samples of core conductor fibre based fabrics in order to determine differences and similarities with regards to various electrostatic properties. This work will be used to establish a new work item proposal within WG-2, Garments, in the ESD Association Standards Committee in the USA.

  5. Degradation of shape memory effect

    International Nuclear Information System (INIS)

    Vandermeer, R.A.

    1983-01-01

    An important parameter for deciding whether or not a SME alloy is suitable for practical applications is the magnitude of the strain reversal accompanying martensite reversion. This research is concerned with elucidating metallurgical factors that cause degradation of this heat-activated recovery strain, E/sub R/. After explaining what is meant by degradation, two manifestations of degradation recently identified in near-monotectoid uranium-niobium alloys are described. The first was associated with the onset of plastic deformation of the martensite beyond the reversible strain limit, E/sub L/; a reduction of E/sub R/ from 5.25% at 8% total strain, i.e. E/sub L/, to 2.9% at 12% total strain was observed. A second type of degradation depended strongly on the heating rate during reversion; the E/sub R/ for an imposed strain of 6.95% was reduced from a value of 5.25% to 1.3% when the heating rate was decreased from 40 0 /sec to 0.05 0 /sec. Degradation was attributed to a change in the transformation path and the interjection of time-dependent, low temperature aging reactions

  6. Um mundo de cores

    Directory of Open Access Journals (Sweden)

    Elis Artz

    2016-08-01

    Full Text Available A pintura de Elis Artz é feita com muita alma e transborda alegria. A vitalidade de seu trabalho transparece nas cores fortes e nos traços simples e harmoniosos. Confira o trabalho da artista nesta edição da Revista Jangada. ELIS by ELIS Descobri meu talento artístisco e criativo há uns 25 anos. Nasci no Brasil e me mudei para os EUA 10 anos atrás por puro amor. Embora seja psicóloga de formação, o meu apreço pela pintura só cresceu e, com o passar dos anos, a paixão pelas tintas me direcionou a fazer cursos com artistas brasileiros renomados. Já morando nos EUA e com essa grande paixão adormecida, durante anos, decidi me entregar para as cores que sempre me trouxeram alegria e cor para os meus dias. Embora muitas de minhas pinturas tenham ido para minha família e amigos no Brasil, vendi inúmeras outras pelo país através de exposições em galerias de arte. Em 2014, fui uma das artistas em destaque no MTD ART nos Estados Unidos. Minha obra estava dentro de cada ônibus das cidades de Champaign e Urbana e exposta em destaque na Estação de Trem. Em maio de 2015, tive o prazer de ter outro trabalho meu nos outdoors da cidade, destacando a minha tela 'Frida' o ano inteiro e de expor em conjunto com alguns artistas locais no final de outubro. Desde então, tenho pintado cada vez mais e me interessado em divulgar o meu trabalho. E, como diria um amigo meu "Elis, você me mostrou que a vida não é só preto no branco". Ele estava certo.

  7. Growth outside the core.

    Science.gov (United States)

    Zook, Chris; Allen, James

    2003-12-01

    Growth in an adjacent market is tougher than it looks; three-quarters of the time, the effort fails. But companies can change those odds dramatically. Results from a five-year study of corporate growth conducted by Bain & Company reveal that adjacency expansion succeeds only when built around strong core businesses that have the potential to become market leaders. And the best place to look for adjacency opportunities is inside a company's strongest customers. The study also found that the most successful companies were able to consistently, profitably outgrow their rivals by developing a formula for pushing out the boundaries of their core businesses in predictable, repeatable ways. Companies use their repeatability formulas to expand into any number of adjacencies. Some companies make repeated geographic moves, as Vodafone has done in expanding from one geographic market to another over the past 13 years, building revenues from $1 billion in 1990 to $48 billion in 2003. Others apply a superior business model to new segments. Dell, for example, has repeatedly adapted its direct-to-customer model to new customer segments and new product categories. In other cases, companies develop hybrid approaches. Nike executed a series of different types of adjacency moves: it expanded into adjacent customer segments, introduced new products, developed new distribution channels, and then moved into adjacent geographic markets. The successful repeaters in the study had two common characteristics. First, they were extraordinarily disciplined, applying rigorous screens before they made an adjacency move. This discipline paid off in the form of learning curve benefits, increased speed, and lower complexity. And second, in almost all cases, they developed their repeatable formulas by studying their customers and their customers' economics very, very carefully.

  8. Core catcher for nuclear reactor core meltdown containment

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Bowman, F.L.

    1978-01-01

    A bed of graphite particles is placed beneath a nuclear reactor core outside the pressure vessel but within the containment building to catch the core debris in the event of failure of the emergency core cooling system. Spray cooling of the debris and graphite particles together with draining and flooding of coolant fluid of the graphite bed is provided to prevent debris slump-through to the bottom of the bed

  9. Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II cold leg injection tests

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

    1985-07-01

    Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase in a PWR-LOCA. In order to investigate the effects of radial power profile, three cold leg injection tests with different radial power profiles under the same total heating power and core stored energy were performed by using the Slab Core Test Facility (SCTF) Core-II. It was revealed by comparing these three tests that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. The turnaround temperature in the high power bundles were evaluated to be reduced by about 40 to 120 K. On the other hand, a two-dimensional flow in the core was also induced by the non-uniform water accumulation in the upper plenum and the quench was delayed resultantly in the bundles corresponding to the peripheral bundles of a PWR. However, the effect of the non-uniform upper plenum water accumulation on the turnaround temperature was small because the effect dominated after the turnaround of the cladding temperature. Selected data from Tests S2-SH1, S2-SH2 and S2-O6 are also presented in this report. Some data from Tests S2-SH1 and S2-SH2 were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

  10. Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

    International Nuclear Information System (INIS)

    Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young; Sang, Seok Yoon

    2014-01-01

    Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

  11. Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Sang, Seok Yoon [Engineering and Technical Center, Korea Hydro, Daejeon (Korea, Republic of)

    2014-08-15

    Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

  12. Comparison of the fractional power motor with cores made of various magnetic materials

    Directory of Open Access Journals (Sweden)

    Gmyrek Zbigniew

    2017-12-01

    Full Text Available The optimization of the motor cores, coupled with new core shapes as well as powering the motor at high frequency are the primary reasons for the use of new materials. The utilization of new materials, like SMC (soft magnetic composite, reduce the core loss and/or provide quasi-isotropic core’s properties in any magnetization direction. Moreover, the use of SMC materials allows for avoiding degradation of the material portions, resulting from punching process, thereby preventing the deterioration of operating parameters of the motor. The authors examine the impact of technological parameters on the properties of a new type of SMC material and analyze the possibility of its use as the core of the fractional power motor. The result of the work is an indication of the shape of the rotor core made of a new SMC material to achieve operational parameters similar to those that have a motor with a core made of laminations.

  13. Comparison of the behaviour of two core designs for ASTRID in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, F., E-mail: frederic.bertrand@cea.fr [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Marie, N.; Prulhière, G.; Lecerf, J. [CEA, DEN, DER, F-13108 Saint Paul-lez-Durance (France); Seiler, J.M. [CEA, DEN, DTN, F-38054 Grenoble (France)

    2016-02-15

    Highlights: • Low void worth CFV and SFRv2 cores are compared for ASTRID pre-conceptual design. • Severe accident behaviour is assessed with a simplified calculation approach and tools. • Mitigation to limit reactivity inserted by core compaction is easier for CFV than for SFRv2 core. • When facing arbitrary reactivity ramps, CFV core would lead to lower energy release than SFRv2 core. • Time scale for core degradation is one order of magnitude larger for CFV than for SFRv2. - Abstract: The present paper is dedicated to the studies carried out during the first stage of the pre-conceptual design of the French demonstrator of fourth generation SFR reactors (ASTRID) in order to compare the behaviour of two envisaged core concepts under severe accident transients. Among the two studied core concepts, whose powers are 1500 MWth, the first one is a classical homogeneous core (called SFRv2) with large pin diameter whose the sodium overall voiding reactivity effect is 5 $. The second concept is an axially heterogeneous core (called CFV) whose global void reactivity effect is negative (−1.2 $ at the end of cycle at the equilibrium). The comparison of the cores relies on two typical accident families: a reactivity insertion (unprotected transient overpower, UTOP) and an overall loss of core cooling (unprotected loss of flow, ULOF). In the first part of the comparison, the primary phase of an UTOP is studied in order to assess typical features of the transient behaviour: power and reactivity evolutions, material heating and melting/vaporization and mechanical energy release due to fuel vapor expansion. The second part of the comparison deals with the calculation of the reactivity potential for degraded states (molten pools) representative of the secondary phase of a mild UTOP and of a strong UTOP (strong or mild qualifies the reactivity ramp inserted). According to the reactivity potential, the amount of fuel to extract from the core and the amount of absorber

  14. Clad Degradation - FEPs Screening Arguments

    International Nuclear Information System (INIS)

    E. Siegmann

    2004-01-01

    The purpose of this report is to document the screening of the cladding degradation features, events, and processes (FEPs) for commercial spent nuclear fuel (CSNF). This report also addresses the effect of some FEPs on both the cladding and the CSNF, DSNF, and HLW waste forms where it was considered appropriate to address the effects on both materials together. This report summarizes the work of others to screen clad degradation FEPs in a manner consistent with, and used in, the Total System Performance Assessment-License Application (TSPA-LA). This document was prepared according to ''Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA'' (BSC 2004a [DIRS 167796])

  15. Radiation degradation of silk protein

    International Nuclear Information System (INIS)

    Pewlong, W.; Sudatis, B.; Takeshita, Hidefumi; Yoshii, Fumio; Kume, Tamikazu

    2000-01-01

    Silk fibroin fiber from the domesticated silkworm Bombyx mori was irradiated using an electron beam accelerator to investigate the application of the radiation degradation technique as a means to solubilize fibroin. The irradiation caused a significant degradation of the fiber. The tensile strength of fibroin fiber irradiated up to 2500 kGy decreased rapidly with increasing dose. The presence of oxygen in the irradiation atmosphere enhanced degradation of the tensile strength. The solubilization of irradiated fibroin fiber was evaluated using the following three kinds of solutions: a calcium chloride solution(CaCl 2 /C 2 H 5 OH/H 2 O=1:2:8 in mole ratio), a hydrochloric acid (0.5 N) and a distilled water. Dissolution of fibroin fiber into these solutions was significantly enhanced by irradiation. Especially, an appreciable amount of water soluble proteins was extracted by a distilled water. (author)

  16. Radiation degradation of silk protein

    Energy Technology Data Exchange (ETDEWEB)

    Pewlong, W; Sudatis, B [Office of Atomic Energy for Peace, Bangkok (Thailand); Takeshita, Hidefumi; Yoshii, Fumio; Kume, Tamikazu [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    2000-03-01

    Silk fibroin fiber from the domesticated silkworm Bombyx mori was irradiated using an electron beam accelerator to investigate the application of the radiation degradation technique as a means to solubilize fibroin. The irradiation caused a significant degradation of the fiber. The tensile strength of fibroin fiber irradiated up to 2500 kGy decreased rapidly with increasing dose. The presence of oxygen in the irradiation atmosphere enhanced degradation of the tensile strength. The solubilization of irradiated fibroin fiber was evaluated using the following three kinds of solutions: a calcium chloride solution(CaCl{sub 2}/C{sub 2}H{sub 5}OH/H{sub 2}O=1:2:8 in mole ratio), a hydrochloric acid (0.5 N) and a distilled water. Dissolution of fibroin fiber into these solutions was significantly enhanced by irradiation. Especially, an appreciable amount of water soluble proteins was extracted by a distilled water. (author)

  17. The transport and behaviour of isoproturon in unsaturated chalk cores

    Science.gov (United States)

    Besien, T. J.; Williams, R. J.; Johnson, A. C.

    2000-04-01

    A batch sorption study, a microcosm degradation study, and two separate column leaching studies were used to investigate the transport and fate of isoproturon in unsaturated chalk. The column leaching studies used undisturbed core material obtained from the field by dry percussion drilling. Each column leaching study used 25 cm long, 10 cm wide unsaturated chalk cores through which a pulse of isoproturon and bromide was eluted. The cores were set-up to simulate conditions in the unsaturated zone of the UK Chalk aquifer by applying a suction of 1 kPa (0.1 m H 2O) to the base of each column, and eluting at a rate corresponding to an average recharge rate through the unsaturated Chalk. A dye tracer indicated that the flow was through the matrix under these conditions. The results from the first column study showed high recovery rates for both isoproturon (73-92%) and bromide (93-96%), and that isoproturon was retarded by a factor of about 1.23 relative to bromide. In the second column study, two of the four columns were eluted with non-sterile groundwater in place of the sterile groundwater used on all other columns, and this study showed high recovery rates for bromide (85-92%) and lower recovery rates for isoproturon (66-79% — sterile groundwater, 48-61% — non-sterile groundwater). The enhanced degradation in the columns eluted with non-sterile groundwater indicated that groundwater microorganisms had increased the degradation rate within these columns. Overall, the reduced isoproturon recovery in the second column study was attributed to increased microbial degradation as a result of the longer study duration (162 vs. 105 days). The breakthrough curves (BTCs) for bromide had a characteristic convection-dispersion shape and were accurately simulated with the minimum of calibration using a simple convection-dispersion model (LEACHP). However, the isoproturon BTCs had an unusual shape and could not be accurately simulated.

  18. Estimation of material degradation of VVER-1000 baffle

    Science.gov (United States)

    Harutyunyan, Davit; Koš'ál, Michal; Vandlík, Stanislav; Hojná, Anna; Schulc, Martin; Flibor, Stanislav

    2017-09-01

    The planned lifetime of the first commercial VVER-1000 units were designed for 30 to 35 years. Most of the early VVER plants are now reaching and/or passing the 35-year mark. Service life extension for another 10 to 30 years is now under investigation. Life extension requires the evaluation of pressure vessel internals degradation under long-term irradiation. One of the possible limiting factors for the service life of VVERs is a void swelling of the Russian type titanium stabilized stainless 08Ch18N10T steel used to construct the baffle surrounding the core. This article aims to show first steps towards deeper analysis of the baffle degradation process and to demonstrate the possibilities of precise calculation and measurements on the VVER-1000 mock-up in LR-0 reactor.

  19. AGR core safety assessment methodologies

    International Nuclear Information System (INIS)

    McLachlan, N.; Reed, J.; Metcalfe, M.P.

    1996-01-01

    To demonstrate the safety of its gas-cooled graphite-moderated AGR reactors, nuclear safety assessments of the cores are based upon a methodology which demonstrates no component failures, geometrical stability of the structure and material properties bounded by a database. All AGRs continue to meet these three criteria. However, predictions of future core behaviour indicate that the safety case methodology will eventually need to be modified to deal with new phenomena. A new approach to the safety assessment of the cores is currently under development, which can take account of these factors while at the same time providing the same level of protection for the cores. This approach will be based on the functionality of the core: unhindered movement of control rods, continued adequate cooling of the fuel and the core, continued ability to charge and discharge fuel. (author). 5 figs

  20. Efficient Test Application for Core-Based Systems Using Twisted-Ring Counters

    OpenAIRE

    Anshuman Chandra; Krishnendu Chakrabarty; Mark C. Hansen

    2001-01-01

    We present novel test set encoding and pattern decompression methods for core-based systems. These are based on the use of twisted-ring counters and offer a number of important advantages–significant test compression (over 10X in many cases), less tester memory and reduced testing time, the ability to use a slow tester without compromising test quality or testing time, and no performance degradation for the core under test. Surprisingly, the encoded test sets obtained from partially-specified...

  1. The Science of Battery Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, John P. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Materials Physics; El Gabaly Marquez, Farid [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Materials Physics; McCarty, Kevin [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Materials Physics; Sugar, Joshua Daniel [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Materials Physics; Talin, Alec A. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Materials Physics; Fenton, Kyle R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Power Sources Design and Development; Nagasubramanian, Ganesan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Power Sources Design and Development; Harris, Charles Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nanosystems Synthesis/Analysis; Jungjohann, Katherine Leigh [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nanosystems Synthesis/Analysis; Hayden, Carl C. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Combustion Chemistry Dept.; Kliewer, Christopher Jesse [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Combustion Chemistry Dept.; Hudak, Nicholas S. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Power Sources Research and Development; Leung, Kevin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Nanostructure Physics; McDaniel, Anthony H. [Sandia National Lab. (SNL-CA), Livermore, CA (United States). Hydrogen and Combustion Technology; Tenney, Craig M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Chemical and Biological Systems; Zavadil, Kevin R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Advanced Materials Lab.

    2015-01-01

    This report documents work that was performed under the Laboratory Directed Research and Development project, Science of Battery Degradation. The focus of this work was on the creation of new experimental and theoretical approaches to understand atomistic mechanisms of degradation in battery electrodes that result in loss of electrical energy storage capacity. Several unique approaches were developed during the course of the project, including the invention of a technique based on ultramicrotoming to cross-section commercial scale battery electrodes, the demonstration of scanning transmission x-ray microscopy (STXM) to probe lithium transport mechanisms within Li-ion battery electrodes, the creation of in-situ liquid cells to observe electrochemical reactions in real-time using both transmission electron microscopy (TEM) and STXM, the creation of an in-situ optical cell utilizing Raman spectroscopy and the application of the cell for analyzing redox flow batteries, the invention of an approach for performing ab initio simulation of electrochemical reactions under potential control and its application for the study of electrolyte degradation, and the development of an electrochemical entropy technique combined with x-ray based structural measurements for understanding origins of battery degradation. These approaches led to a number of scientific discoveries. Using STXM we learned that lithium iron phosphate battery cathodes display unexpected behavior during lithiation wherein lithium transport is controlled by nucleation of a lithiated phase, leading to high heterogeneity in lithium content at each particle and a surprising invariance of local current density with the overall electrode charging current. We discovered using in-situ transmission electron microscopy that there is a size limit to lithiation of silicon anode particles above which particle fracture controls electrode degradation. From electrochemical entropy measurements, we discovered that entropy

  2. A comparative genomic analysis of the oxidative enzymes potentially involved in lignin degradation by Agaricus bisporus

    Science.gov (United States)

    Harshavardhan Doddapaneni; Venkataramanan Subramanian; Bolei Fu; Dan Cullen

    2013-01-01

    The oxidative enzymatic machinery for degradation of organic substrates in Agaricus bisporus (Ab) is at the core of the carbon recycling mechanisms in this fungus. To date, 156 genes have been tentatively identified as part of this oxidative enzymatic machinery, which includes 26 peroxidase encoding genes, nine copper radical oxidase [including three...

  3. Construction of the HTTR in-core components

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Jinza, K.; Miki, T.

    1996-01-01

    The reactor internals of HTTR consist of graphite and metallic core support structures and shielding blocks and are designed to support core elements and to shield neutron fluence. They also have functions to restrict by-pass flow for ensuring the core cooling performance and to maintain the temperature of metallic core support structures within their design limits. The detailed design of the HTTR core support structure was approved by the government through safety review, 1990-1991. Machining of all graphite components, which consist of about 150 large blocks, was finished in September 1994 successfully. Machining and fabricating of the metallic components were also finished in September. Prior to their installation in the reactor pressure vessel (RPV), the assembly test of actual reactor internals was performed at the works to confirm above mentioned functions. The assembly test was conducted by examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the RPV and reactor internals as well as under the core support plates with respect to structural integrity, and measuring by-pass flow rate through gaps between graphite components which may degrade core performance. The another purpose of the assembly test was to confirm the installation procedure of those components. All components were assembled at the works according to the planned procedure, and the tests were executed while assembling. As a result of the tests, measured level difference and gap width between reactor internals were negligible from core thermal and hydraulic performance point of view. Coolant flows uniformly in circumferential direction at any axial level in the RPV. By-pass flow rate was found to be suppressed sufficiently and far less than the design limit. (J.P.N.)

  4. Overview of core disruptive accidents

    International Nuclear Information System (INIS)

    Marchaterre, J.F.

    1977-01-01

    An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

  5. Coupling between core and cladding modes in a helical core fiber with large core offset

    International Nuclear Information System (INIS)

    Napiorkowski, Maciej; Urbanczyk, Waclaw

    2016-01-01

    We analyzed the effect of resonant coupling between core and cladding modes in a helical core fiber with large core offset using the fully vectorial method based on the transformation optics formalism. Our study revealed that the resonant couplings to lower order cladding modes predicted by perturbative methods and observed experimentally in fibers with small core offsets are in fact prohibited for larger core offsets. This effect is related to the lack of phase matching caused by elongation of the optical path of the fundamental modes in the helical core. Moreover, strong couplings to the cladding modes of the azimuthal modal number much higher than predicted by perturbative methods may be observed for large core offsets, as the core offset introduces higher order angular harmonics in the field distribution of the fundamental modes. Finally, in contrast to previous studies, we demonstrate the existence of spectrally broad polarization sensitive couplings to the cladding modes suggesting that helical core fibers with large core offsets may be used as broadband circular polarizers. (paper)

  6. A simplified geometrical model for transient corium propagation in core for LWR with heavy reflector - 15271

    International Nuclear Information System (INIS)

    Saas, L.; Le Tellier, R.; Bajard, S.

    2015-01-01

    In this document, we present a simplified geometrical model (0D model) for both the in-core corium propagation transient and the characterization of the mode of corium transfer from the core to the vessel. A degraded core with a formed corium pool is used as an initial state. This initial state can be obtained from a simulation computed with an integral code. This model does not use a grid for the core as integral codes do. Geometrical shapes and 0D models are associated with the corium pool and the other components of the degraded core (debris, heavy reflector, core plate...). During the transient, these shapes evolve taking into account the thermal and stratification behavior of the corium pool and the melting of the core surrounding components. Some results corresponding to the corium pool propagation in core transients obtained with this model on a LWR with a heavy reflector are given and compared to grid approach of the integral codes MAAP4

  7. Models of the earth's core

    International Nuclear Information System (INIS)

    Stevenson, D.J.

    1981-01-01

    The combination of seismology, high pressure experiment and theory, geomagnetism, fluid dynamics, and current views of terrestrial planetary evolution lead to strong constraints on core models. The synthesis presented here is devoted to the defense of the following properties: (1) core formation was contemporaneous with earth accretion; (2) the outer, liquid core is predominately iron but cannot be purely iron; (3) the inner core-outer core boundary represents a thermodynamic equilibrium between a liquid alloys and a predominanately iron solid; (4) thermodynamic and transport properties of outer core can be estimated from liquid-state theories; and (5) the outer core is probably adiabatic and uniform in composition. None of these propositions are universally accepted by geophysicists. But, the intent of this paper is to present a coherent picture which explains most of the data with the fewest ad hoc assumptions. Areas in which future progress is both essential and likely are geo- and cosmochronology, seismological determinations of core structure, fluid dynamics of the core and mantle, and condensed matter physics

  8. Wire core reactor for NTP

    International Nuclear Information System (INIS)

    Harty, R.B.

    1991-01-01

    The development of the wire core system for Nuclear Thermal Propulsion (NTP) that took place from 1963 to 1965 is discussed. A wire core consists of a fuel wire with spacer wires. It's an annular flow core having a central control rod. There are actually four of these, with beryllium solid reflectors on both ends and all the way around. Much of the information on the concept is given in viewgraph form. Viewgraphs are presented on design details of the wire core, the engine design, engine weight vs. thrust, a technique used to fabricate the wire fuel element, and axial temperature distribution

  9. Nuclear reactor with several cores

    International Nuclear Information System (INIS)

    Swars, H.

    1977-01-01

    Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

  10. Biochemistry Instrumentation Core Technology Center

    Data.gov (United States)

    Federal Laboratory Consortium — The UCLA-DOE Biochemistry Instrumentation Core Facility provides the UCLA biochemistry community with easy access to sophisticated instrumentation for a wide variety...

  11. Core barrel inner tube lifter

    Energy Technology Data Exchange (ETDEWEB)

    Jeffers, J P

    1968-07-16

    A core drill with means for selectively lifting a core barrel inner tube consists of a lifting means connected to the core barrel inner tube assembly. It has a closable passage to permit drilling fluid normally to pass through it. The lifting means has a normally downward facing surface and a means to direct drilling fluid pressure against that surface so that on closure of the passage to fluid flow, the pressure of the drilling fluid is caused to act selectively on it. This causes the lifting means to rise and lift the core barrel. (7 claims)

  12. TMI-2 core debris analysis

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  13. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  14. Fort St. Vrain core performance

    International Nuclear Information System (INIS)

    McEachern, D.W.; Brown, J.R.; Heller, R.A.; Franek, W.J.

    1977-07-01

    The Fort St. Vrain High Temperature Gas Cooled Reactor core performance has been evaluated during the startup testing phase of the reactor operation. The reactor is graphite moderated, helium cooled, and uses coated particle fuel and on-line flow control to each of the 37 refueling regions. Principal objectives of startup testing were to determine: core and control system reactivity, radial power distribution, flow control capability, and initial fission product release. Information from the core demonstrates that Technical Specifications are being met, performance of the core and fuel is as expected, flow and reactivity control are predictable and simple for the operator to carry out

  15. Characterizing the Core via K-Core Covers

    NARCIS (Netherlands)

    Sanchez, S.M.; Borm, P.E.M.; Estevez, A.

    2013-01-01

    This paper extends the notion of individual minimal rights for a transferable utility game (TU-game) to coalitional minimal rights using minimal balanced families of a specific type, thus defining a corresponding minimal rights game. It is shown that the core of a TU-game coincides with the core of

  16. Adult educators' core competences

    Science.gov (United States)

    Wahlgren, Bjarne

    2016-06-01

    Which competences do professional adult educators need? This research note discusses the topic from a comparative perspective, finding that adult educators' required competences are wide-ranging, heterogeneous and complex. They are subject to context in terms of national and cultural environment as well as the kind of adult education concerned (e.g. basic education, work-related education etc.). However, it seems that it is possible to identify certain competence requirements which transcend national, cultural and functional boundaries. This research note summarises these common or "core" requirements, organising them into four thematic subcategories: (1) communicating subject knowledge; (2) taking students' prior learning into account; (3) supporting a learning environment; and (4) the adult educator's reflection on his or her own performance. At the end of his analysis of different competence profiles, the author notes that adult educators' ability to train adult learners in a way which then enables them to apply and use what they have learned in practice (thus performing knowledge transfer) still seems to be overlooked.

  17. Emergency core cooling system

    International Nuclear Information System (INIS)

    Arai, Kenji; Oikawa, Hirohide.

    1990-01-01

    The device according to this invention can ensure cooling water required for emerency core cooling upon emergence such as abnormally, for example, loss of coolant accident, without using dynamic equipments such as a centrifugal pump or large-scaled tank. The device comprises a pressure accumulation tank containing a high pressure nitrogen gas and cooling water inside, a condensate storage tank, a pressure suppression pool and a jet stream pump. In this device there are disposed a pipeline for guiding cooling water in the pressure accumulation tank as a jetting water to a jetting stream pump, a pipeline for guiding cooling water stored in the condensate storage tank and the pressure suppression pool as pumped water to the jetting pump and, further, a pipeline for guiding the discharged water from the jet stream pump which is a mixed stream of pumped water and jetting water into the reactor pressure vessel. In this constitution, a sufficient amount of water ranging from relatively high pressure to low pressure can be supplied into the reactor pressure vessel, without increasing the size of the pressure accumulation tank. (I.S.)

  18. Emergency core cooling system

    International Nuclear Information System (INIS)

    Sato, Akira; Kobayashi, Masahide.

    1983-01-01

    Purpose: To enable a stable operation of an emergency core cooling system by preventing the system from the automatic stopping at an abnormally high level of the reactor water during its operation. Constitution: A pump flow rate signal and a reactor water level signal are used and, when the reactor water level is increased to a predetermined level, the pump flow rate is controlled by the reactor water level signal instead of the flow rate signal. Specifically, when the reactor water level is gradually increased by the water injection from the pump and exceeds a setting signal for the water level, the water level deviation signal acts as a demand signal for the decrease in the flow rate of the pump and the output signal from the water level controller is also decreased depending on the control constant. At a certain point, the output signal from the water level controller becomes smaller than the output signal from the flow rate controller. Thus, the output signal from the water level controller is outputted as the output signal for the lower level preference device. In this way, the reactor water level and the pump flow rate can be controlled within a range not exceeding the predetermined pump flow rate. (Horiuchi, T.)

  19. Emergency core cooling system

    International Nuclear Information System (INIS)

    Abe, Nobuaki.

    1993-01-01

    A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

  20. Necrosome core machinery: MLKL.

    Science.gov (United States)

    Zhang, Jing; Yang, Yu; He, Wenyan; Sun, Liming

    2016-06-01

    In the study of regulated cell death, the rapidly expanding field of regulated necrosis, in particular necroptosis, has been drawing much attention. The signaling of necroptosis represents a sophisticated form of a death pathway. Anti-caspase mechanisms (e.g., using inhibitors of caspases, or genetic ablation of caspase-8) switch cell fate from apoptosis to necroptosis. The initial extracellular death signals regulate RIP1 and RIP3 kinase activation. The RIP3-associated death complex assembly is necessary and sufficient to initiate necroptosis. MLKL was initially identified as an essential mediator of RIP1/RIP3 kinase-initiated necroptosis. Recent studies on the signal transduction using chemical tools and biomarkers support the idea that MLKL is able to make more functional sense for the core machinery of the necroptosis death complex, called the necrosome, to connect to the necroptosis execution. The experimental data available now have pointed that the activated MLKL forms membrane-disrupting pores causing membrane leakage, which extends the prototypical concept of morphological and biochemical events following necroptosis happening in vivo. The key role of MLKL in necroptosis signaling thus sheds light on the logic underlying this unique "membrane-explosive" cell death pathway. In this review, we provide the general concepts and strategies that underlie signal transduction of this form of cell death, and then focus specifically on the role of MLKL in necroptosis.

  1. Preliminary analysis for u tube degradation in CANDU steam generator using CATHENA

    Energy Technology Data Exchange (ETDEWEB)

    Shin, So Eun; Lee, Jeong Hun; Park, Tong Kyu; Hwang, Su Hyun [FNC Technology Co., Seoul (Korea, Republic of); Jung, Jong Yeo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The interest in plant safety and integrity has been increasing due to long term operation of nuclear power plants (NPPs) and lots of efforts have been devoted to developing the degradation evaluation model for all the Structure, System, and Components (SSCs) of NPPs in these days. The efforts, however, were mainly concentrated on pressurized light water reactors (PWRs) in domestic. In contrast, the study for the aging degradation of counterparts of CANDU (CANada Deuterium Uranium) reactors has been rarely performed, even though Wolsong unit 1 (WS1), that is a CANDU 6 NPP in Korea, has been operating for almost 30 years. Therefore, the assessment of the aging degradation is required and the proper and exact evaluation model for the aging degradation of SCCs of CANDU, especially WS1, is urgently needed. In this study, the aging degradation of steam generators (SGs) in WS1 was mainly discussed. Based on cases of the aging degradation of SGs in overseas CANDU reactors, the major potential aging mechanisms of SGs were estimated since there has been no case of accident due to degradation in CANDU NPPs in Korea . Some core parameters which are indicators of the degree of degradation were calculated by CATHENA (Canadian algorithm for thermal hydraulic network analysis). In the result of comparing two calculation cases; core parameters for only aged SGs in fresh plant and those for all the aged component, it can be concluded that aging of SGs is a main component in the degradation assessment of CANDU NPPs, and keeping the integrity of steam generator (SG) tubes is important to guarantee the safety of the NPPs.

  2. Degradation of materials and passivity

    International Nuclear Information System (INIS)

    Meisel, W.

    1997-01-01

    Demanding for a reduction in materials degradation is a serious problem all over the world. Moessbauer spectroscopy (MS) is, among others, a very valuable tool to follow many degradation processes. Evidently, Fe is the most important Moessbauer element considering the overall presence of iron in everyday life. MS may contribute to our knowledge about nearly all fields of materials degradation, chemical, mechanical, thermal, irradiative, etc. Following some general lines, corrosion is considered in particular. MS is applicable to investigate the bulk of materials as well as their surface layers with an information depth of ca. 250 nm. In general, it has to be applied as a surface sensitive method in combination with other relevant methods in order to get a detailed insight into ongoing processes. Some examples have been selected to elucidate the application of MS in this field. Another class of examples concerns attempts to prevent corrosion, i.e., the application of coatings and transforming chemicals. A very effective and most natural way to reduce corrosion is the passivation of materials. The effect of passive layers and their destruction by environmental influences are discussed using results of MS and related methods. It is outlined that passivity is not restricted to chemically treated metals but can be considered as a general concept for preventing different kinds of materials from degradation. (orig.)

  3. Land degradation and property regimes

    Science.gov (United States)

    Paul M. Beaumont; Robert T. Walker

    1996-01-01

    This paper addresses the relationship between property regimes and land degradation outcomes, in the context of peasant agriculture. We consider explicitly whether private property provides for superior soil resource conservation, as compared to common property and open access. To assess this we implement optimization algorithms on a supercomputer to address resource...

  4. Degradation of CIGS solar cells

    NARCIS (Netherlands)

    Theelen, M.J.

    2015-01-01

    Large scale commercial introduction of CIGS photovoltaics (PV) requires modules with low costs, high efficiencies and long and predictable lifetimes. Unfortunately,knowledge about the lifetime of CIGS PV is limited, which is reflected in the results of field studies: degradation rates varying from

  5. The Degradation of a Nation.

    Science.gov (United States)

    Morozova, Galina Fedorouna

    1995-01-01

    Maintains that the process of national degradation is a real danger and concern of all Russian society. Discusses environmental concerns, such as water, soil, and air pollution; falling birth rates; aging of the population; crime; and decline in moral values. Concludes that it is imperative for all citizens to stop and reverse these trends. (CFR)

  6. Polymeric Materials - introduction and degradation

    DEFF Research Database (Denmark)

    Kontogeorgis, Georgios

    1999-01-01

    These notes support the polymer part of the courses 91742 and 91762 (Materials and Corrosion/degradation of materials) taught in IFAKthey contain a short introduction on group contribution methods for estimating properties of polymers, polymer thermodynamics, viscoelasticity models as well...

  7. Abiotic degradation of antibiotic ionophores

    DEFF Research Database (Denmark)

    Bohn, Pernille; Bak, Søren A; Björklund, Erland

    2013-01-01

    Hydrolytic and photolytic degradation were investigated for the ionophore antibiotics lasalocid, monensin, salinomycin, and narasin. The hydrolysis study was carried out by dissolving the ionophores in solutions of pH 4, 7, and 9, followed by incubation at three temperatures of 6, 22, and 28 °C f...... because they absorb light of environmentally irrelevant wavelengths....

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  9. Core-to-core uniformity improvement in multi-core fiber Bragg gratings

    Science.gov (United States)

    Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

    2014-07-01

    Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

  10. Thermo-mechanical interaction effects in foam cored sandwich panels-correlation between High-order models and Finite element analysis results

    DEFF Research Database (Denmark)

    Palleti, Hara Naga Krishna Teja; Santiuste, Carlos; Thomsen, Ole Thybo

    2010-01-01

    Thermo-mechanical interaction effects including thermal material degradation in polymer foam cored sandwich structures is investigated using the commercial Finite Element Analysis (FEA) package ABAQUS/Standard. Sandwich panels with different boundary conditions in the form of simply supported...

  11. Refurbishment, core conversion and safety analysis of Apsara reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K.; Sasidharan, K.; Sengupta, S. [Bhabha Atomic Research Centre, Mumbai (India)]. E-mail: nram@@apsara.barc.ernet.in

    1998-07-01

    Apsara, a 1 MWt pool type reactor using HEU fuel has been in operation at the Bhabha Atomic Research Centre, Trombay since 1956. In view of the long service period seen by the reactor it is now planned to carry out extensive refurbishment of the reactor with a view to extend its useful life. It is also proposed to modify the design of the reactor wherein the core will be surrounded by a heavy water reflector tank to obtain a good thermal neutron flux over a large radial distance from the core. Beam holes and the majority of the irradiation facilities will be located inside the reflector tank. The coolant flow direction through the core will be changed from the existing upward flow to downward flow. A delay tank, located inside the pool, is provided to facilitate decay of short lived radioactivity in the coolant outlet from the core in order to bring down radiation field in the operating areas. Analysis of various anticipated operational occurrences and accident conditions like loss of normal power, core coolant flow bypass, fuel channel blockage and degradation of primary coolant pressure boundary have been performed for the proposed design. Details of the proposed design modifications and the safety analyses are given in the paper. (author)

  12. Nuclear reactor core stabilizing arrangement

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1976-01-01

    A nuclear reactor core stabilizing arrangement is described wherein a plurality of actuators, disposed in a pattern laterally surrounding a group of elongated fuel assemblies, press against respective contiguous fuel assemblies on the periphery of the group to reduce the clearance between adjacent fuel assemblies thereby forming a more compacted, vibration resistant core structure. 7 claims, 4 drawing figures

  13. Complicated Politics to the Core

    Science.gov (United States)

    McGuinn, Patrick

    2015-01-01

    People dislike the Common Core for several different reasons, and so it is important to disaggregate the sources of opposition and to assess and then to dispel some of the myths that have built up around it. It also is important to understand the unusual political alliances that have emerged in opposition to Common Core implementation and how they…

  14. Toward full MOX core design

    International Nuclear Information System (INIS)

    Rouviere, G.; Guillet, J.L.; Bruna, G.B.; Pelet, J.

    1999-01-01

    This paper presents a selection of the main preliminary results of a study program sponsored by COGEMA and currently carried out by FRAMATOME. The objective of this study is to investigate the feasibility of full MOX core loading in a French 1300 MWe PWR, a recent and widespread standard nuclear power plant. The investigation includes core nuclear design, thermal hydraulic and systems aspects. (authors)

  15. Collapsing stellar cores and supernovae

    Energy Technology Data Exchange (ETDEWEB)

    Epstein, R J [Nordisk Inst. for Teoretisk Atomfysik, Copenhagen (Denmark); Noorgaard, H [Nordisk Inst. for Teoretisk Atomfysik, Copenhagen (Denmark); Chicago Univ., IL (USA). Enrico Fermi Inst.); Bond, J R [Niels Bohr Institutet, Copenhagen (Denmark); California Inst. of Tech., Pasadena (USA). W.K. Kellogg Radiation Lab.)

    1979-05-01

    The evolution of a stellar core is studied during its final quasi-hydrostatic contraction. The core structure and the (poorly known) properties of neutron rich matter are parametrized to include most plausible cases. It is found that the density-temperature trajectory of the material in the central part of the core (the core-center) is insensitive to nearly all reasonable parameter variations. The central density at the onset of the dynamic phase of the collapse (when the core-center begins to fall away from the rest of the star) and the fraction of the emitted neutrinos which are trapped in the collapsing core-center depend quite sensitively on the properties of neutron rich matter. We estimate that the amount of energy Ecm which is imparted to the core-mantle by the neutrinos which escape from the imploded core-center can span a large range of values. For plausible choices of nuclear and model parameters Ecm can be large enough to yield a supernova event.

  16. The INTEGRAL Core Observing Programme

    DEFF Research Database (Denmark)

    Winkler, C.; Gehrels, N.; Lund, Niels

    1999-01-01

    The Core Programme of the INTEGRAL mission is defined as the portion of the scientific programme covering the guaranteed time observations for the INTEGRAL Science Working Team. This paper describes the current status of the Core Programme preparations and summarizes the key elements...... of the observing programme....

  17. Core barrel plug

    International Nuclear Information System (INIS)

    Tolino, R.W.; Hopkins, R.J.; Congleton, R.L.; Popalis, C.H.

    1986-01-01

    A plug is described for preventing flow through a port in a core barrel of a pressurized water nuclear reactor which consists of: a substantially cylindrical body formed with a cylindrical portion and a flange and defining a tapered leading open end with the other end being closed by an end plug attached to the flange, the body defining a bore therein extending from the open end to the end plug with the bore having a smaller diameter near the open end than near the end plug, the cylindrical portion having a lip near the open end and being formed with longitudinal slots extending from the open end toward the flange and extending entirely through the thickness of the cylindrical portion, the cylindrical portion having a circumferential first groove on the outer surface thereof located near the forwardmost portion of the cylindrical portion but not in the section of the cylindrical portion that has the slots therein, and a plurality of circumferential second grooves on the outer surface thereof located in the section of the cylindrical portion that has the slots therein, the first and second grooves establishing a seal between the cylindrical portion and the inside surface of the port when the cylindrical portion is expanded, and the flange and the end plug having a passageway defined therein for introducing a fluid into the body; a metal ring disposed in each of the second grooves; a mandrel slidably disposed and captured in the body and capable of being moved toward the open end of the body when the fluid is introduced through the passageway, thereby causing the cylindrical portion to be expanded into contact with the inside surface of the port; and a locking mechanism disposed in the end plug for preventing inadvertent movement of the mandrel

  18. Radiation resistivity of pure-silica core image guide

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. SCORPIO - VVER core surveillance system

    International Nuclear Information System (INIS)

    Zalesky, K.; Svarny, J.; Novak, L.; Rosol, J.; Horanes, A.

    1997-01-01

    The Halden Project has developed the core surveillance system SCORPIO which has two parallel modes of operation: the Core Follow Mode and the Predictive Mode. The main motivation behind the development of SCORPIO is to make a practical tool for reactor operators which can increase the quality and quantity of information presented on core status and dynamic behavior. This can first of all improve plant safety as undesired core conditions are detected and prevented. Secondly, more flexible and efficient plant operation is made possible. So far the system has only been implemented on western PWRs but the basic concept is applicable to a wide range of reactor including WWERs. The main differences between WWERs and typical western PWRs with respect to core surveillance requirements are outlined. The development of a WWER version of SCORPIO was initiated in cooperation with the Nuclear Research Institute at Rez and industry partners in the Czech Republic. The first system will be installed at the Dukovany NPP. (author)

  20. Core body temperature in obesity.

    Science.gov (United States)

    Heikens, Marc J; Gorbach, Alexander M; Eden, Henry S; Savastano, David M; Chen, Kong Y; Skarulis, Monica C; Yanovski, Jack A

    2011-05-01

    A lower core body temperature set point has been suggested to be a factor that could potentially predispose humans to develop obesity. We tested the hypothesis that obese individuals have lower core temperatures than those in normal-weight individuals. In study 1, nonobese [body mass index (BMI; in kg/m(2)) temperature-sensing capsules, and we measured core temperatures continuously for 24 h. In study 2, normal-weight (BMI of 18-25) and obese subjects swallowed temperature-sensing capsules to measure core temperatures continuously for ≥48 h and kept activity logs. We constructed daily, 24-h core temperature profiles for analysis. Mean (±SE) daily core body temperature did not differ significantly between the 35 nonobese and 46 obese subjects (36.92 ± 0.03°C compared with 36.89 ± 0.03°C; P = 0.44). Core temperature 24-h profiles did not differ significantly between 11 normal-weight and 19 obese subjects (P = 0.274). Women had a mean core body temperature ≈0.23°C greater than that of men (36.99 ± 0.03°C compared with 36.76 ± 0.03°C; P body temperature. It may be necessary to study individuals with function-altering mutations in core temperature-regulating genes to determine whether differences in the core body temperature set point affect the regulation of human body weight. These trials were registered at clinicaltrials.gov as NCT00428987 and NCT00266500.

  1. Reactor Core Internals Replacement of Ikata Units 1 and 2

    International Nuclear Information System (INIS)

    Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

    2012-01-01

    This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

  2. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  3. Pu recycling in a full Th-MOX PWR core. Part I: Steady state analysis

    International Nuclear Information System (INIS)

    Fridman, E.; Kliem, S.

    2011-01-01

    Research highlights: → Detailed 3D 100% Th-MOX PWR core design is developed. → Pu incineration increased by a factor of 2 as compared to a full MOX PWR core. → The core controllability under steady state conditions is demonstrated. - Abstract: Current practice of Pu recycling in existing Light Water Reactors (LWRs) in the form of U-Pu mixed oxide fuel (MOX) is not efficient due to continuous Pu production from U-238. The use of Th-Pu mixed oxide (TOX) fuel will considerably improve Pu consumption rates because virtually no new Pu is generated from thorium. In this study, the feasibility of Pu recycling in a typical pressurized water reactor (PWR) fully loaded with TOX fuel is investigated. Detailed 3-dimensional 100% TOX and 100% MOX PWR core designs are developed. The full MOX core is considered for comparison purposes. The design stages included determination of Pu loading required to achieve 18-month fuel cycle assuming three-batch fuel management scheme, selection of poison materials, development of the core loading pattern, optimization of burnable poison loadings, evaluation of critical boron concentration requirements, estimation of reactivity coefficients, core kinetic parameters, and shutdown margin. The performance of the MOX and TOX cores under steady-state condition and during selected reactivity initiated accidents (RIAs) is compared with that of the actual uranium oxide (UOX) PWR core. Part I of this paper describes the full TOX and MOX PWR core designs and reports the results of steady state analysis. The TOX core requires a slightly higher initial Pu loading than the MOX core to achieve the target fuel cycle length. However, the TOX core exhibits superior Pu incineration capabilities. The significantly degraded worth of control materials in Pu cores is partially addressed by the use of enriched soluble boron and B 4 C as a control rod absorbing material. Wet annular burnable absorber (WABA) rods are used to flatten radial power distribution

  4. Two-dimensional thermal-hydraulic behavior in core in SCTF Core-II forced feed reflood tests

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Sobajima, Makoto; Okubo, Tsutomu; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

    1987-01-01

    Major purpose of the Slab Core Test Program is to investigate the two-dimensional thermal-hydraulic behavior in the core during the reflood phase of a PWR-LOCA. It was revealed in the previous Slab Core Test Facility (SCTF) Core-II test results that the heat transfer was enhanced in the higher power bundles and degraded in the lower power bundles in the non-uniform radial power profile tests. In order to separately evaluate the effect of the radial power (Q) distribution itself and the effect of the radial temperature (T) distribution, four tests were performed with steep Q and T, flat Q and T, steep Q and flat T, and flat Q and steep T. Based on the test results, it was concluded that the radial temperature distribution which accompanied the radial power distribution was the dominant factor of the two-dimensional thermal-hydraulic behavior in the core during the initial period. Selected data from these four tests are also presented in this report. Some data from Test S2-12 (steep Q, T) were compared with TRAC post-test calculations performed by the Los Alamos National Laboratory. (author)

  5. Epistemology and ontology in core ontologies: FOLaw and LRI-Core, two core ontologies for law

    NARCIS (Netherlands)

    Breukers, J.A.P.J.; Hoekstra, R.J.

    2004-01-01

    For more than a decade constructing ontologies for legal domains, we, at the Leibniz Center for Law, felt really the need to develop a core ontology for law that would enable us to re-use the common denominator of the various legal domains. In this paper we present two core ontologies for law. The

  6. Degradation in steam of 60 cm-long B{sub 4}C control rods

    Energy Technology Data Exchange (ETDEWEB)

    Dominguez, C., E-mail: christina.dominguez@irsn.fr; Drouan, D.

    2014-08-01

    In the framework of nuclear reactor core meltdown accident studies, the degradation of boron carbide control rod segments exposed to argon/steam atmospheres was investigated up to about 2000 °C in IRSN laboratories. The sequence of the phenomena involved in the degradation has been found to take place as expected. Nevertheless, the ZrO{sub 2} oxide layer formed on the outer surface of the guide tube was very protective, significantly delaying and limiting the guide tube failure and therefore the boron carbide pellet oxidation. Contrary to what was expected, the presence of the control rod decreases the hydrogen release instead of increasing it by additional oxidation of boron compounds. Boron contents up to 20 wt.% were measured in metallic mixtures formed during degradation. It was observed that these metallic melts are able to attack the surrounding fuel rods, which could have consequences on fuel degradation and fission product release kinetics during severe accidents.

  7. Advanced Oxidation Degradation of Diclofenac

    Energy Technology Data Exchange (ETDEWEB)

    Cooper, William J., E-mail: wcooper@uci.edu [Urban Water Research Center, Department of Civil and Environmental Engineering, University of California, Irvine, CA 92697 (United States); Song Weihua, E-mail: wsong@fudan.edu.cn [Department of Environmental Science & Engineering, Fudan University, Shanghai 200433 (China)

    2012-07-01

    Advanced oxidation/reduction processes (AO/RPs), utilize free radical reactions to directly degrade chemical contaminants as an alternative to traditional water treatment. This study reports the absolute rate constants for reaction of diclofenac sodium and the model compound (2, 6-dichloraniline) with the two major AO/RP radicals; the hydroxyl radical (•OH) and hydrated electron (e{sup -}{sub aq}). The bimolecular reaction rate constants (M{sup -1} s{sup -1}) for diclofenac for •OH was (9.29 ± 0.11) x 10{sup 9}, and, for e- aq was (1.53 ± 0.03) x10{sup 9}. Preliminary degradation mechanisms are suggested based on product analysis using {sup 60}Co γ-irradiation and LC-MS for reaction by-product identification. The toxicity of products was evaluated using the Vibrio fischeri luminescent bacteria method. (author)

  8. Advanced Oxidation Degradation of Diclofenac

    International Nuclear Information System (INIS)

    Cooper, William J.; Song Weihua

    2012-01-01

    Advanced oxidation/reduction processes (AO/RPs), utilize free radical reactions to directly degrade chemical contaminants as an alternative to traditional water treatment. This study reports the absolute rate constants for reaction of diclofenac sodium and the model compound (2, 6-dichloraniline) with the two major AO/RP radicals; the hydroxyl radical (•OH) and hydrated electron (e - aq ). The bimolecular reaction rate constants (M -1 s -1 ) for diclofenac for •OH was (9.29 ± 0.11) x 10 9 , and, for e- aq was (1.53 ± 0.03) x10 9 . Preliminary degradation mechanisms are suggested based on product analysis using 60 Co γ-irradiation and LC-MS for reaction by-product identification. The toxicity of products was evaluated using the Vibrio fischeri luminescent bacteria method. (author)

  9. Fungal degradation of organophosphorous insecticides

    Energy Technology Data Exchange (ETDEWEB)

    Bumpus, J.A. [Notre Dame Univ., IN (United States); Kakar, S.N.; Coleman, R.D. [Argonne National Lab., IL (United States)

    1992-07-01

    Organophosphorous insecticides are used extensively to treat a variety of pests and insects. Although as a group they are easily degraded by bacteria in the environment, a number of them have half-lives of several months. Little is known about their biodegradation by fungi. We have shown that Phanerochaete chrysosporium can substantially degrade chlorpyrifos, fonofos, and terbufos (27.5%, 12.2%, and 26.6%, respectively) during 18-day incubation in nitrogen-limited stationary cultures. The results demonstrate that the clorinated pyridinyl ring of chlorpyrifos and the phenyl ring of fonofos undergo ring cleavage during biodegradation by the fungus. The usefulness of the fungus system for bioremediation is discussed. 16 refs., 7 figs., 2 tabs.

  10. Fungal degradation of organophosphorous insecticides

    Energy Technology Data Exchange (ETDEWEB)

    Bumpus, J.A. (Notre Dame Univ., IN (United States)); Kakar, S.N.; Coleman, R.D. (Argonne National Lab., IL (United States))

    1992-01-01

    Organophosphorous insecticides are used extensively to treat a variety of pests and insects. Although as a group they are easily degraded by bacteria in the environment, a number of them have half-lives of several months. Little is known about their biodegradation by fungi. We have shown that Phanerochaete chrysosporium can substantially degrade chlorpyrifos, fonofos, and terbufos (27.5%, 12.2%, and 26.6%, respectively) during 18-day incubation in nitrogen-limited stationary cultures. The results demonstrate that the clorinated pyridinyl ring of chlorpyrifos and the phenyl ring of fonofos undergo ring cleavage during biodegradation by the fungus. The usefulness of the fungus system for bioremediation is discussed. 16 refs., 7 figs., 2 tabs.

  11. Radiation degradation of silk protein

    Energy Technology Data Exchange (ETDEWEB)

    Wachiraporn Pewlong; Boonya Sudatis [Office of Atomic Energy for Peace, Bangkok (Thailand); Takeshita, Hidefumi; Yoshii, Fumio; Kume, Tamikazu [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    2000-09-01

    Silk fibroin fiber from the domesticated silkworm Bombyx mori was irradiated in the dose range up to 2500 kGy using an electron beam accelerator to apply the radiation degradation technique as a means to solubilize fibroin. The tensile strength of irradiated fibroin fiber decreased with increasing dose and the presence of oxygen in the irradiation atmosphere enhanced the degradation. The solubilization of irradiated fibroin fiber was evaluated using the following three kinds of solutions: calcium chloride solution (CaCl{sub 2}/C{sub 2}H{sub 5}OH/H{sub 2}O = 1 : 2 : 8 in mole ratio), hydrochloric acid (0.5N) and distilled water. Dissolution of fibroin fiber into these solutions was significantly enhanced by irradiation. Especially, an appreciable amount of water-soluble protein was extracted by distilled water. (author)

  12. HOW STARLESS ARE STARLESS CORES?

    International Nuclear Information System (INIS)

    Schnee, Scott; Friesen, Rachel; Di Francesco, James; Johnstone, Doug; Enoch, Melissa; Sadavoy, Sarah

    2012-01-01

    In this paper, we present the results of Combined Array for Research in Millimeter-wave Astronomy continuum and spectral line observations of the dense core Per-Bolo 45. Although this core has previously been classified as starless, we find evidence for an outflow and conclude that Per-Bolo 45 is actually an embedded, low-luminosity protostar. We discuss the impact of newly discovered, low-luminosity, embedded objects in the Perseus molecular cloud on starless core and protostar lifetimes. We estimate that the starless core lifetime has been overestimated by 4%-18% and the Class 0/I protostellar lifetime has been underestimated by 5%-20%. Given the relatively large systematic uncertainties involved in these calculations, variations on the order of 10% do not significantly change either core lifetimes or the expected protostellar luminosity function. Finally, we suggest that high-resolution (sub)millimeter surveys of known cores lacking near-infrared and mid-infrared emission are necessary to make an accurate census of starless cores.

  13. Challenges for proteomics core facilities.

    Science.gov (United States)

    Lilley, Kathryn S; Deery, Michael J; Gatto, Laurent

    2011-03-01

    Many analytical techniques have been executed by core facilities established within academic, pharmaceutical and other industrial institutions. The centralization of such facilities ensures a level of expertise and hardware which often cannot be supported by individual laboratories. The establishment of a core facility thus makes the technology available for multiple researchers in the same institution. Often, the services within the core facility are also opened out to researchers from other institutions, frequently with a fee being levied for the service provided. In the 1990s, with the onset of the age of genomics, there was an abundance of DNA analysis facilities, many of which have since disappeared from institutions and are now available through commercial sources. Ten years on, as proteomics was beginning to be utilized by many researchers, this technology found itself an ideal candidate for being placed within a core facility. We discuss what in our view are the daily challenges of proteomics core facilities. We also examine the potential unmet needs of the proteomics core facility that may also be applicable to proteomics laboratories which do not function as core facilities. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. Degradation and inhibition of cyclooxygenase

    OpenAIRE

    Neuß, Heiko

    2011-01-01

    The cyclooxygenase (COX) is a central enzyme in the genesis of pain, inflammation and carcinogenesis. Two major isoforms, COX-1 and COX-2, have been described. The COX-1 is constitutively expressed in most tissues and has housekeeping functions, whereas the COX-2 is the inducible isoform, expressed under conditions of inflammation and tumor growth. First, we researched the degradation of the COX-2 enzyme. We were able to demonstrate, that the COX-2 protein was ubiquitinated before prote...

  15. Core principles of evolutionary medicine

    Science.gov (United States)

    Grunspan, Daniel Z; Nesse, Randolph M; Barnes, M Elizabeth; Brownell, Sara E

    2018-01-01

    Abstract Background and objectives Evolutionary medicine is a rapidly growing field that uses the principles of evolutionary biology to better understand, prevent and treat disease, and that uses studies of disease to advance basic knowledge in evolutionary biology. Over-arching principles of evolutionary medicine have been described in publications, but our study is the first to systematically elicit core principles from a diverse panel of experts in evolutionary medicine. These principles should be useful to advance recent recommendations made by The Association of American Medical Colleges and the Howard Hughes Medical Institute to make evolutionary thinking a core competency for pre-medical education. Methodology The Delphi method was used to elicit and validate a list of core principles for evolutionary medicine. The study included four surveys administered in sequence to 56 expert panelists. The initial open-ended survey created a list of possible core principles; the three subsequent surveys winnowed the list and assessed the accuracy and importance of each principle. Results Fourteen core principles elicited at least 80% of the panelists to agree or strongly agree that they were important core principles for evolutionary medicine. These principles over-lapped with concepts discussed in other articles discussing key concepts in evolutionary medicine. Conclusions and implications This set of core principles will be helpful for researchers and instructors in evolutionary medicine. We recommend that evolutionary medicine instructors use the list of core principles to construct learning goals. Evolutionary medicine is a young field, so this list of core principles will likely change as the field develops further. PMID:29493660

  16. Single gene retrieval from thermally degraded DNA

    Indian Academy of Sciences (India)

    Unknown

    DNA thermal degradation was shown to occur via a singlet oxygen pathway. A comparative study of the ther- mal degradation of cellular DNA and isolated DNA showed that cellular ..... definite level of energy (e.g. depurination active energy,.

  17. Extensions and applications of degradation modeling

    International Nuclear Information System (INIS)

    Hsu, F.; Subudhi, M.; Samanta, P.K.; Vesely, W.E.

    1991-01-01

    Component degradation modeling being developed to understand the aging process can have many applications with potential advantages. Previous work has focused on developing the basic concepts and mathematical development of a simple degradation model. Using this simple model, times of degradations and failures occurrences were analyzed for standby components to detect indications of aging and to infer the effectiveness of maintenance in preventing age-related degradations from transforming to failures. Degradation modeling approaches can have broader applications in aging studies and in this paper, the authors discuss some of the extensions and applications of degradation modeling. The extensions and applications of the degradation modeling approaches discussed are: (a) theoretical developments to study reliability effects of different maintenance strategies and policies, (b) relating aging-failure rate to degradation rate, and (c) application to a continuously operating component

  18. Changes in Flow and Transport Patterns in Fen Peat as a Result of Soil Degradation

    Science.gov (United States)

    Liu, Haojie; Janssen, Manon; Lennartz, Bernd

    2016-04-01

    The preferential movement of water and transport of substances play an important role in soils and are not yet fully understood especially in degraded peat soils. In this study, we aimed at deducing changes in flow and transport patterns in the course of soil degradation as resulting from peat drainage, using titanium dioxide (TiO2) as a dye tracer. The dye tracer experiments were conducted on columns of eight types of differently degraded peat soils from three sites taken both in vertical and horizontal directions. The titanium dioxide suspension (average particle size of 0.3 μm; 10 g l-1) was applied in a pulse of 40 mm to each soil core. Twenty-four hours after the application of the tracer, cross sections of the soil cores were prepared for photo documentation. In addition, the saturated hydraulic conductivity (Ks) was determined. Preferential flow occurred in all investigated peat types. From the stained soil structural elements, we concluded that undecomposed plant remains are the major preferential flow pathways in less degraded peat. For more strongly degraded peat, bio-pores, such as root and earthworm channels, operated as the major transport domain. Results show that Ks and the effective pore network in less degraded peat soils are anisotropic. With increasing peat degradation, the Ks and cross section of effective pore network decreased. The results also indicate a strong positive relationship between Ks and number of macropores as well as pore continuity. Hence, we conclude that changes in flow and transport pathways as well as Ks with an increasing peat degradation are due to the disintegration of the peat forming plant material and decrement of number and continuity of macropores after drainage.

  19. Life extension of CANDU reactor cores

    International Nuclear Information System (INIS)

    Millard, J.; Kerker, J.; Albert, M.

    2011-01-01

    Candu Energy (formerly AECL), in partnership with station operators, has developed a robust methodology for demonstrating the fitness of reactor core structures, and associated reactivity control devices, as an essential element in conducting a station life extension project. The ageing of reactors is affected by ageing mechanisms impacted by operational history and design related factors such as materials, chemistries and stress distributions. The methodology of this life extension work is based on the IAEA TECDOC 1197; which documents practices for ageing management in CANDU reactors. This paper uses the work in Bruce Units 1 and 2, conducted from 2007 through to 2011, to explain the methodology. The work started with analysis of historical operational conditions and identification of the forms of degradation that could have occurred. The assessment and related inspections considered the safety and pressure boundary significance of each item, as well as its failure modes and margins. It then moved through both general and local inspection, focused mainly inside the calandria vessel once the calandria tubes were removed. The inspection found the bulk of the hardware to be in good condition, with a small number of remediation opportunities. In the course of that remediation some foreign material was sampled and removed. The minor remediation was successful and the work was completed through formal documentation of the fitness for extended life. It has been demonstrated through these analyses and visual inspections that the reactor structures and components inspected are free of indications and active degradation mechanisms that would prevent the safe and reliable operation of Bruce A Units 1 and 2 through its next 25 years of life. (author)

  20. Diagnostic Technology Development for Core Internal Structure in CANDU reactor

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

    2005-04-01

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  1. Assessing Core Competencies

    Science.gov (United States)

    Narayanan, M.

    2004-12-01

    Catherine Palomba and Trudy Banta offer the following definition of assessment, adapted from one provided by Marches in 1987. Assessment in the systematic collection, review, and use of information about educational programs undertaken for the purpose of improving student learning and development. (Palomba and Banta 1999). It is widely recognized that sophisticated computing technologies are becoming a key element in today's classroom instructional techniques. Regardless, the Professor must be held responsible for creating an instructional environment in which the technology actually supplements learning outcomes of the students. Almost all academic disciplines have found a niche for computer-based instruction in their respective professional domain. In many cases, it is viewed as an essential and integral part of the educational process. Educational institutions are committing substantial resources to the establishment of dedicated technology-based laboratories, so that they will be able to accommodate and fulfill students' desire to master certain of these specific skills. This type of technology-based instruction may raise some fundamental questions about the core competencies of the student learner. Some of the most important questions are : 1. Is the utilization of these fast high-powered computers and user-friendly software programs creating a totally non-challenging instructional environment for the student learner ? 2. Can technology itself all too easily overshadow the learning outcomes intended ? 3. Are the educational institutions simply training students how to use technology rather than educating them in the appropriate field ? 4. Are we still teaching content-driven courses and analysis oriented subject matter ? 5. Are these sophisticated modern era technologies contributing to a decline in the Critical Thinking Capabilities of the 21st century technology-savvy students ? The author tries to focus on technology as a tool and not on the technology

  2. Flow accelerated organic coating degradation

    Science.gov (United States)

    Zhou, Qixin

    Applying organic coatings is a common and the most cost effective way to protect metallic objects and structures from corrosion. Water entry into coating-metal interface is usually the main cause for the deterioration of organic coatings, which leads to coating delamination and underfilm corrosion. Recently, flowing fluids over sample surface have received attention due to their capability to accelerate material degradation. A plethora of works has focused on the flow induced metal corrosion, while few studies have investigated the flow accelerated organic coating degradation. Flowing fluids above coating surface affect corrosion by enhancing the water transport and abrading the surface due to fluid shear. Hence, it is of great importance to understand the influence of flowing fluids on the degradation of corrosion protective organic coatings. In this study, a pigmented marine coating and several clear coatings were exposed to the laminar flow and stationary immersion. The laminar flow was pressure driven and confined in a flow channel. A 3.5 wt% sodium chloride solution and pure water was employed as the working fluid with a variety of flow rates. The corrosion protective properties of organic coatings were monitored inline by Electrochemical Impedance Spectroscopy (EIS) measurement. Equivalent circuit models were employed to interpret the EIS spectra. The time evolution of coating resistance and capacitance obtained from the model was studied to demonstrate the coating degradation. Thickness, gloss, and other topography characterizations were conducted to facilitate the assessment of the corrosion. The working fluids were characterized by Fourier Transform Infrared Spectrometer (FTIR) and conductivity measurement. The influence of flow rate, fluid shear, fluid composition, and other effects in the coating degradation were investigated. We conclude that flowing fluid on the coating surface accelerates the transport of water, oxygen, and ions into the coating, as

  3. Applications and extensions of degradation modeling

    International Nuclear Information System (INIS)

    Hsu, F.; Subudhi, M.; Samanta, P.K.; Vesely, W.E.

    1991-01-01

    Component degradation modeling being developed to understand the aging process can have many applications with potential advantages. Previous work has focused on developing the basic concepts and mathematical development of a simple degradation model. Using this simple model, times of degradations and failures occurrences were analyzed for standby components to detect indications of aging and to infer the effectiveness of maintenance in preventing age-related degradations from transforming to failures. Degradation modeling approaches can have broader applications in aging studies and in this paper, we discuss some of the extensions and applications of degradation modeling. The application and extension of degradation modeling approaches, presented in this paper, cover two aspects: (1) application to a continuously operating component, and (2) extension of the approach to analyze degradation-failure rate relationship. The application of the modeling approach to a continuously operating component (namely, air compressors) shows the usefulness of this approach in studying aging effects and the role of maintenance in this type component. In this case, aging effects in air compressors are demonstrated by the increase in both the degradation and failure rate and the faster increase in the failure rate compared to the degradation rate shows the ineffectiveness of the existing maintenance practices. Degradation-failure rate relationship was analyzed using data from residual heat removal system pumps. A simple linear model with a time-lag between these two parameters was studied. The application in this case showed a time-lag of 2 years for degradations to affect failure occurrences. 2 refs

  4. Degradation analysis of thin film photovoltaic modules

    International Nuclear Information System (INIS)

    Radue, C.; Dyk, E.E. van

    2009-01-01

    Five thin film photovoltaic modules were deployed outdoors under open circuit conditions after a thorough indoor evaluation. Two technology types were investigated: amorphous silicon (a-Si:H) and copper indium gallium diselenide (CIGS). Two 14 W a-Si:H modules, labelled Si-1 and Si-2, were investigated. Both exhibited degradation, initially due to the well-known light-induced degradation described by Staebler and Wronski [Applied Physics Letters 31 (4) (1977) 292], and thereafter due to other degradation modes such as cell degradation. The various degradation modes contributing to the degradation of the a-Si:H modules will be discussed. The initial maximum power output (P MAX ) of Si-1 was 9.92 W, with the initial light-induced degradation for Si-1 ∼30% and a total degradation of ∼42%. For Si-2 the initial P MAX was 7.93 W, with initial light-induced degradation of ∼10% and a total degradation of ∼17%. Three CIGS modules were investigated: two 20 W modules labelled CIGS-1 and CIGS-2, and a 40 W module labelled CIGS-3. CIGS-2 exhibited stable performance while CIGS-1 and CIGS-3 exhibited degradation. CIGS is known to be stable over long periods of time, and thus the possible reasons for the degradation of the two modules are discussed.

  5. Degradation analysis of thin film photovoltaic modules

    Energy Technology Data Exchange (ETDEWEB)

    Radue, C., E-mail: chantelle.radue@nmmu.ac.z [Department of Physics, PO Box 77000, Nelson Mandela Metropolitan University, Port Elizabeth 6031 (South Africa); Dyk, E.E. van [Department of Physics, PO Box 77000, Nelson Mandela Metropolitan University, Port Elizabeth 6031 (South Africa)

    2009-12-01

    Five thin film photovoltaic modules were deployed outdoors under open circuit conditions after a thorough indoor evaluation. Two technology types were investigated: amorphous silicon (a-Si:H) and copper indium gallium diselenide (CIGS). Two 14 W a-Si:H modules, labelled Si-1 and Si-2, were investigated. Both exhibited degradation, initially due to the well-known light-induced degradation described by Staebler and Wronski [Applied Physics Letters 31 (4) (1977) 292], and thereafter due to other degradation modes such as cell degradation. The various degradation modes contributing to the degradation of the a-Si:H modules will be discussed. The initial maximum power output (P{sub MAX}) of Si-1 was 9.92 W, with the initial light-induced degradation for Si-1 approx30% and a total degradation of approx42%. For Si-2 the initial P{sub MAX} was 7.93 W, with initial light-induced degradation of approx10% and a total degradation of approx17%. Three CIGS modules were investigated: two 20 W modules labelled CIGS-1 and CIGS-2, and a 40 W module labelled CIGS-3. CIGS-2 exhibited stable performance while CIGS-1 and CIGS-3 exhibited degradation. CIGS is known to be stable over long periods of time, and thus the possible reasons for the degradation of the two modules are discussed.

  6. Applications and extensions of degradation modeling

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, F.; Subudhi, M.; Samanta, P.K. [Brookhaven National Lab., Upton, NY (United States); Vesely, W.E. [Science Applications International Corp., Columbus, OH (United States)

    1991-12-31

    Component degradation modeling being developed to understand the aging process can have many applications with potential advantages. Previous work has focused on developing the basic concepts and mathematical development of a simple degradation model. Using this simple model, times of degradations and failures occurrences were analyzed for standby components to detect indications of aging and to infer the effectiveness of maintenance in preventing age-related degradations from transforming to failures. Degradation modeling approaches can have broader applications in aging studies and in this paper, we discuss some of the extensions and applications of degradation modeling. The application and extension of degradation modeling approaches, presented in this paper, cover two aspects: (1) application to a continuously operating component, and (2) extension of the approach to analyze degradation-failure rate relationship. The application of the modeling approach to a continuously operating component (namely, air compressors) shows the usefulness of this approach in studying aging effects and the role of maintenance in this type component. In this case, aging effects in air compressors are demonstrated by the increase in both the degradation and failure rate and the faster increase in the failure rate compared to the degradation rate shows the ineffectiveness of the existing maintenance practices. Degradation-failure rate relationship was analyzed using data from residual heat removal system pumps. A simple linear model with a time-lag between these two parameters was studied. The application in this case showed a time-lag of 2 years for degradations to affect failure occurrences. 2 refs.

  7. Applications and extensions of degradation modeling

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, F.; Subudhi, M.; Samanta, P.K. (Brookhaven National Lab., Upton, NY (United States)); Vesely, W.E. (Science Applications International Corp., Columbus, OH (United States))

    1991-01-01

    Component degradation modeling being developed to understand the aging process can have many applications with potential advantages. Previous work has focused on developing the basic concepts and mathematical development of a simple degradation model. Using this simple model, times of degradations and failures occurrences were analyzed for standby components to detect indications of aging and to infer the effectiveness of maintenance in preventing age-related degradations from transforming to failures. Degradation modeling approaches can have broader applications in aging studies and in this paper, we discuss some of the extensions and applications of degradation modeling. The application and extension of degradation modeling approaches, presented in this paper, cover two aspects: (1) application to a continuously operating component, and (2) extension of the approach to analyze degradation-failure rate relationship. The application of the modeling approach to a continuously operating component (namely, air compressors) shows the usefulness of this approach in studying aging effects and the role of maintenance in this type component. In this case, aging effects in air compressors are demonstrated by the increase in both the degradation and failure rate and the faster increase in the failure rate compared to the degradation rate shows the ineffectiveness of the existing maintenance practices. Degradation-failure rate relationship was analyzed using data from residual heat removal system pumps. A simple linear model with a time-lag between these two parameters was studied. The application in this case showed a time-lag of 2 years for degradations to affect failure occurrences. 2 refs.

  8. Modelling land degradation in IMAGE 2

    NARCIS (Netherlands)

    Hootsmans RM; Bouwman AF; Leemans R; Kreileman GJJ; MNV

    2001-01-01

    Food security may be threatened by loss of soil productivity as a result of human-induced land degradation. Water erosion is the most important cause of land degradation, and its effects are irreversible. This report describes the IMAGE land degradation model developed for describing current and

  9. SpaceCube Core Software

    Data.gov (United States)

    National Aeronautics and Space Administration — Develop a flexible, modular and user friendly SpaceCube Core Software system that will dramatically simplify SpaceCube application development and enable any...

  10. Viral Evolution Core | FNLCR Staging

    Science.gov (United States)

    Brandon F. Keele, Ph.D. PI/Senior Principal Investigator, Retroviral Evolution Section Head, Viral Evolution Core Leidos Biomedical Research, Inc. Frederick National Laboratory for Cancer Research Frederick, MD 21702-1201 Tel: 301-846-173

  11. Computed tomography of drill cores

    International Nuclear Information System (INIS)

    Taylor, T.

    1985-08-01

    A preliminary computed tomography evaluation of drill cores of granite and sandstone has generated geologically significant data. Density variations as small as 4 percent and fractures as narrow as 0.1 mm were easily detected

  12. In-core monitoring detectors

    International Nuclear Information System (INIS)

    Mitelman, M.G.

    2001-01-01

    The main task of in-core monitoring consists in securing observability of the reactor installation in all possible operation modes (normal, transient, accident and post-accident). Operation safety at acceptable cost can be achieved by optimized measurement errors. The range of sensors applied as in-core detectors for operative measurements in the industry is very limited in number. Among them might be cited self powered neutron detectors (SPND) and thermocouples. Sensors are incorporated in the in-core detectors assemblies (SVRD). The presentation makes an effort to touch upon the problems of assuring and increasing quality of in-core on-line measurements. So we do not consider systems using movable detectors, as the latter do not assure on-line measurements. (Authors)

  13. Discovering the Army's Core Competencies

    National Research Council Canada - National Science Library

    Rudesheim, Frederick

    2001-01-01

    This paper seeks to answer the question, "Has the Army correctly identified its core competencies to ensure the Army can adequately respond to the national military strategy?" FM 1, The Army (Prototype Draft...

  14. Structural degradation of Thar lignite using MW1 fungal isolate: optimization studies

    Science.gov (United States)

    Haider, Rizwan; Ghauri, Muhammad A.; Jones, Elizabeth J.; Orem, William H.; SanFilipo, John R.

    2015-01-01

    Biological degradation of low-rank coals, particularly degradation mediated by fungi, can play an important role in helping us to utilize neglected lignite resources for both fuel and non-fuel applications. Fungal degradation of low-rank coals has already been investigated for the extraction of soil-conditioning agents and the substrates, which could be subjected to subsequent processing for the generation of alternative fuel options, like methane. However, to achieve an efficient degradation process, the fungal isolates must originate from an appropriate coal environment and the degradation process must be optimized. With this in mind, a representative sample from the Thar coalfield (the largest lignite resource of Pakistan) was treated with a fungal strain, MW1, which was previously isolated from a drilled core coal sample. The treatment caused the liberation of organic fractions from the structural matrix of coal. Fungal degradation was optimized, and it showed significant release of organics, with 0.1% glucose concentration and 1% coal loading ratio after an incubation time of 7 days. Analytical investigations revealed the release of complex organic moieties, pertaining to polyaromatic hydrocarbons, and it also helped in predicting structural units present within structure of coal. Such isolates, with enhanced degradation capabilities, can definitely help in exploiting the chemical-feedstock-status of coal.

  15. Fungal degradation of coal as a pretreatment for methane production

    Science.gov (United States)

    Haider, Rizwan; Ghauri, Muhammad A.; SanFilipo, John R.; Jones, Elizabeth J.; Orem, William H.; Tatu, Calin A.; Akhtar, Kalsoom; Akhtar, Nasrin

    2013-01-01

    Coal conversion technologies can help in taking advantage of huge low rank coal reserves by converting those into alternative fuels like methane. In this regard, fungal degradation of coal can serve as a pretreatment step in order to make coal a suitable substrate for biological beneficiation. A fungal isolate MW1, identified as Penicillium chrysogenum on the basis of fungal ITS sequences, was isolated from a core sample of coal, taken from a well drilled by the US. Geological Survey in Montana, USA. The low rank coal samples, from major coal fields of Pakistan, were treated with MW1 for 7 days in the presence of 0.1% ammonium sulfate as nitrogen source and 0.1% glucose as a supplemental carbon source. Liquid extracts were analyzed through Excitation–Emission Matrix Spectroscopy (EEMS) to obtain qualitative estimates of solubilized coal; these analyses indicated the release of complex organic functionalities. In addition, GC–MS analysis of these extracts confirmed the presence of single ring aromatics, polyaromatic hydrocarbons (PAHs), aromatic nitrogen compounds and aliphatics. Subsequently, the released organics were subjected to a bioassay for the generation of methane which conferred the potential application of fungal degradation as pretreatment. Additionally, fungal-mediated degradation was also prospected for extracting some other chemical entities like humic acids from brown coals with high huminite content especially from Thar, the largest lignite reserve of Pakistan.

  16. Recriticality analyses for CAPRA cores

    International Nuclear Information System (INIS)

    Maschek, W.; Thiem, D.

    1995-01-01

    The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

  17. Recriticality analyses for CAPRA cores

    Energy Technology Data Exchange (ETDEWEB)

    Maschek, W.; Thiem, D.

    1995-08-01

    The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

  18. One dimensional reactor core model

    International Nuclear Information System (INIS)

    Kostadinov, V.; Stritar, A.; Radovo, M.; Mavko, B.

    1984-01-01

    The one dimensional model of neutron dynamic in reactor core was developed. The core was divided in several axial nodes. The one group neutron diffusion equation for each node is solved. Feedback affects of fuel and water temperatures is calculated. The influence of xenon, boron and control rods is included in cross section calculations for each node. The system of equations is solved implicitly. The model is used in basic principle Training Simulator of NPP Krsko. (author)

  19. Improvements to core-catchers

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, T C.W.

    1969-07-22

    A core catcher consists of a generally annular holder adapted to be contained within a core barrel with sets of dogs circumferentially disposed in the holder. Each set of dogs consists of at least 2 dogs of different lengths pivotally mounted in the holder to swing inward. The dogs in each set are vertically superimposed. They are of upward decreasing length, with the arc of swing of the vertically adjacent dogs overlapping. (8 claims)

  20. Nanoporous polymer liquid core waveguides

    DEFF Research Database (Denmark)

    Gopalakrishnan, Nimi; Christiansen, Mads Brøkner; Ndoni, Sokol

    2010-01-01

    We demonstrate liquid core waveguides defined by UV to enable selective water infiltration in nanoporous polymers, creating an effective refractive index shift Δn=0.13. The mode confinement and propagation loss in these waveguides are presented.......We demonstrate liquid core waveguides defined by UV to enable selective water infiltration in nanoporous polymers, creating an effective refractive index shift Δn=0.13. The mode confinement and propagation loss in these waveguides are presented....

  1. Reactive transport modeling of chemical and isotope data to identify degradation processes of chlorinated ethenes in a diffusion-dominated media

    DEFF Research Database (Denmark)

    Chambon, Julie Claire Claudia; Damgaard, Ida; Jeannottat, Simon

    . Degradation and transport processes of chlorinated ethenes are not well understood in such geological settings, therefore risk assessment and remediation at these sites are particularly challenging. In this work, a combined approach of chemical and isotope analysis on core samples, and reactive transport...... the source zone (between 6 and 12 mbs). Concentrations and stable isotope ratios of the mother compounds and their daughter products, as well as redox parameters, fatty acids and microbial data, were analyzed with discrete sub-sampling along the cores. More samples (each 5 mm) were collected around...... of dechlorination and degradation pathways (biotic reductive dechlorination or abiotic β-elimination with iron minerals) in three core profiles. The model includes diffusion in the matrix, sequential reductive dechlorination, abiotic degradation, isotope fractionation due to degradation and due to diffusion...

  2. All Metal Iron Core For A Low Aspect Ratio Tokamak

    International Nuclear Information System (INIS)

    Gates, D.A.; Jun, C.; Zatz, I.; Zolfaghari, A.

    2010-01-01

    A novel concept for incorporating a iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance separators between the toroidal field windings. This design avoids the inherent problems of a multiturn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron-induced conductivity. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance separators. Useful loop voltage time histories have been obtained using achievable resistivities.

  3. LMFBR Ultra Long Life Cores

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Doncals, R.A.; Porter, C.A.; Gundy, L.M.

    1986-01-01

    The Ultra Long Life Core is an attractive and innovative design approach with several extremely beneficial attributes. Long Life cores are applicable to the full range of LMR plant sizes resulting in lifetimes up to 30 years. Core life is somewhat limited for smaller plant sizes, however significant benefits of this approach still exist for all plant sizes. The union of long life cores and the complementary inherent safety technology offer a means of utilizing the well-proven oxide fuel in a system with unsurpassed safety capability. A further benefit is that the uranium fuel cycle can be used in long life cores, especially for initial LMR plant deployment, thereby eliminating the need for reprocessing prior to starting LMR plant construction in the U.S. Finally the long life core significantly reduces power costs. With inherent safety capability designed into an LMR and with the ULLC fuel cycle, power costs competitive with light water plants are achievable while offering improved operational flexibility derived through extending refueling intervals

  4. TMI-2 core boring machine

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  5. Reactor core performance calculating device

    International Nuclear Information System (INIS)

    Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.

    1995-01-01

    The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)

  6. Core cooling system for reactor

    International Nuclear Information System (INIS)

    Kondo, Ryoichi; Amada, Tatsuo.

    1976-01-01

    Purpose: To improve the function of residual heat dissipation from the reactor core in case of emergency by providing a secondary cooling system flow channel, through which fluid having been subjected to heat exchange with the fluid flowing in a primary cooling system flow channel flows, with a core residual heat removal system in parallel with a main cooling system provided with a steam generator. Constitution: Heat generated in the core during normal reactor operation is transferred from a primary cooling system flow channel to a secondary cooling system flow channel through a main heat exchanger and then transferred through a steam generator to a water-steam system flow channel. In the event if removal of heat from the core by the main cooling system becomes impossible due to such cause as breakage of the duct line of the primary cooling system flow channel or a trouble in a primary cooling system pump, a flow control valve is opened, and steam generator inlet and outlet valves are closed, thus increasing the flow rate in the core residual heat removal system. Thereafter, a blower is started to cause dissipation of the core residual heat from the flow channel of a system for heat dissipation to atmosphere. (Seki, T.)

  7. Assessing and Projecting Greenhouse Gas Release due to Abrupt Permafrost Degradation

    Science.gov (United States)

    Saito, K.; Ohno, H.; Yokohata, T.; Iwahana, G.; Machiya, H.

    2017-12-01

    Permafrost is a large reservoir of frozen soil organic carbon (SOC; about half of all the terrestrial storage). Therefore, its degradation (i.e., thawing) under global warming may lead to a substantial amount of additional greenhouse gas (GHG) release. However, understanding of the processes, geographical distribution of such hazards, and implementation of the relevant processes in the advanced climate models are insufficient yet so that variations in permafrost remains one of the large source of uncertainty in climatic and biogeochemical assessment and projections. Thermokarst, induced by melting of ground ice in ice-rich permafrost, leads to dynamic surface subsidence up to 60 m, which further affects local and regional societies and eco-systems in the Arctic. It can also accelerate a large-scale warming process through a positive feedback between released GHGs (especially methane), atmospheric warming and permafrost degradation. This three-year research project (2-1605, Environment Research and Technology Development Fund of the Ministry of the Environment, Japan) aims to assess and project the impacts of GHG release through dynamic permafrost degradation through in-situ and remote (e.g., satellite and airborn) observations, lab analysis of sampled ice and soil cores, and numerical modeling, by demonstrating the vulnerability distribution and relative impacts between large-scale degradation and such dynamic degradation. Our preliminary laboratory analysis of ice and soil cores sampled in 2016 at the Alaskan and Siberian sites largely underlain by ice-rich permafrost, shows that, although gas volumes trapped in unit mass are more or less homogenous among sites both for ice and soil cores, large variations are found in the methane concentration in the trapped gases, ranging from a few ppm (similar to that of the atmosphere) to hundreds of thousands ppm We will also present our numerical approach to evaluate relative impacts of GHGs released through dynamic

  8. Silicon Nanophotonics for Many-Core On-Chip Networks

    Science.gov (United States)

    Mohamed, Moustafa

    Number of cores in many-core architectures are scaling to unprecedented levels requiring ever increasing communication capacity. Traditionally, architects follow the path of higher throughput at the expense of latency. This trend has evolved into being problematic for performance in many-core architectures. Moreover, the trends of power consumption is increasing with system scaling mandating nontraditional solutions. Nanophotonics can address these problems, offering benefits in the three frontiers of many-core processor design: Latency, bandwidth, and power. Nanophotonics leverage circuit-switching flow control allowing low latency; in addition, the power consumption of optical links is significantly lower compared to their electrical counterparts at intermediate and long links. Finally, through wave division multiplexing, we can keep the high bandwidth trends without sacrificing the throughput. This thesis focuses on realizing nanophotonics for communication in many-core architectures at different design levels considering reliability challenges that our fabrication and measurements reveal. First, we study how to design on-chip networks for low latency, low power, and high bandwidth by exploiting the full potential of nanophotonics. The design process considers device level limitations and capabilities on one hand, and system level demands in terms of power and performance on the other hand. The design involves the choice of devices, designing the optical link, the topology, the arbitration technique, and the routing mechanism. Next, we address the problem of reliability in on-chip networks. Reliability not only degrades performance but can block communication. Hence, we propose a reliability-aware design flow and present a reliability management technique based on this flow to address reliability in the system. In the proposed flow reliability is modeled and analyzed for at the device, architecture, and system level. Our reliability management technique is

  9. Core/shell PLGA microspheres with controllable in vivo release profile via rational core phase design.

    Science.gov (United States)

    Yu, Meiling; Yao, Qing; Zhang, Yan; Chen, Huilin; He, Haibing; Zhang, Yu; Yin, Tian; Tang, Xing; Xu, Hui

    2018-02-27

    the microspheres prepared by various methods were mainly controlled by either the porosity inside the microspheres or the degradation of materials, which could, therefore, lead to different release behaviours. This results indicated great potential of the PLGA microsphere formulation as an injectable depot for controllable in vivo release profile via rational core phase design. Core/shell microspheres fabricated by modified double emulsification-solvent evaporation methods, with various inner phases, to obtain high loading drugs system, as well as appropriate release behaviours. Accordingly, control in vivo release profile via rational core phase design.

  10. Nonlinear seismic analysis of a graphite reactor core

    International Nuclear Information System (INIS)

    Laframboise, W.L.; Desmond, T.P.

    1988-01-01

    Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)

  11. Using microorganisms to aid in hydrocarbon degradation

    International Nuclear Information System (INIS)

    Black, W.; Zamora, J.

    1993-01-01

    Aliphatic hydrocarbons are threatening the potable water supply and the aquatic ecosystem. Given the right microbial inhabitant(s), a large portion of these aliphatic hydrocarbons could be biodegraded before reaching the water supply. The authors' purpose is to isolate possible oil-degrading organisms. Soil samples were taken from hydrocarbon-laden soils at petroleum terminals, a petroleum refinery waste-treatment facility, a sewage-treatment plant grease collector, a site of previous bioremediation, and various other places. Some isolates known to be good degraders were obtained from culture collection services. These samples were plated on a 10w-30 multigrade motor oil solid medium to screen for aliphatic hydrocarbon degraders. The degrading organisms were isolated, identified, and tested (CO 2 evolution, BOD, and COD) to determine the most efficient degrader(s). Thirty-seven organisms were tested, and the most efficient degraders were Serratia marcescens, Escherichia coli, and Enterobacter agglomerans

  12. Enzymatic degradation of polycaprolactone–gelatin blend

    International Nuclear Information System (INIS)

    Banerjee, Aditi; Chatterjee, Kaushik; Madras, Giridhar

    2015-01-01

    Blends of polycaprolactone (PCL), a synthetic polymer and gelatin, natural polymer offer a optimal combination of strength, water wettability and cytocompatibility for use as a resorbable biomaterial. The enzymatic degradation of PCL, gelatin and PCL–gelatin blended films was studied in the presence of lipase (Novozym 435, immobilized) and lysozyme. Novozym 435 degraded the PCL films whereas lysozyme degraded the gelatin. Though Novozym 435 and lysozyme individually could degrade PCL–gelatin blended films, the combination of these enzymes showed the highest degradation of these blended films. Moreover, the enzymatic degradation was much faster when fresh enzymes were added at regular intervals. The changes in physico-chemical properties of polymer films due to degradation were studied by scanning electron microscopy, Fourier transform infrared spectroscopy and differential scanning calorimetry. These results have important implications for designing resorbable biomedical implants. (paper)

  13. Study of PP/montmorillonite composite degradation

    International Nuclear Information System (INIS)

    Baer, Marcia; Granado, Carlos J.F.

    2009-01-01

    The objective of this work was to produce composites of PP/sodium bentonite and PP/ organophilic bentonite through melt intercalation and analyze the degradation produced by ultraviolet irradiation. The XRD results showed that the samples of nature bentonite had better interaction with de polymer and produced intercalated nanocomposite. The effect of UV irradiation on degradation was observed after 24 hours of exposition. The samples showed the same photoproducts and at the same proportion until 240 hours of UV exposition; with 480 hours the organophilize bentonite composite showed higher degradation than other ones. The superficial cracks increased with degradation time. The degradation occurs due chromophores impurities presented in the samples, thus samples with sodium clay show higher degradation, and organophilic clay contains ammonium salt that contribute to increase the degradation. (author)

  14. STRUCTURAL PERFORMANCE OF DEGRADED REINFORCED CONCRETE MEMBERS

    International Nuclear Information System (INIS)

    Braverman, J.I.; Miller, C.A.; Ellingwood, B.R.; Naus, D.J.; Hofmayer, C.H.; Bezler, P.; Chang, T.Y.

    2001-01-01

    This paper describes the results of a study to evaluate, in probabilistic terms, the effects of age-related degradation on the structural performance of reinforced concrete members at nuclear power plants. The paper focuses on degradation of reinforced concrete flexural members and shear walls due to the loss of steel reinforcing area and loss of concrete area (cracking/spalling). Loss of steel area is typically caused by corrosion while cracking and spalling can be caused by corrosion of reinforcing steel, freeze-thaw, or aggressive chemical attack. Structural performance in the presence of uncertainties is depicted by a fragility (or conditional probability of failure). The effects of degradation on the fragility of reinforced concrete members are calculated to assess the potential significance of various levels of degradation. The fragility modeling procedures applied to degraded concrete members can be used to assess the effects of degradation on plant risk and can lead to the development of probability-based degradation acceptance limits

  15. Analysis of BWR/Mark III drywell failure during degraded core accidents

    International Nuclear Information System (INIS)

    Yang, J.W.

    1983-01-01

    The potential for a hydrogen detonation due to the accumulation of a large amount of hydrogen in the drywell region of a BWR Mark III containment is analyzed. Loss of integrity of the drywell wall causes a complete bypass of the suppression pool and leads to pressurization of the containment building. However, the predicted peak containment pressure does not exceed the estimates of containment failure pressure

  16. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

  17. Nuclear plant service water system aging degradation assessment

    International Nuclear Information System (INIS)

    Jarrell, D.B.; Larson, L.L.; Stratton, R.C.; Bohn, S.J.; Gore, M.L.

    1992-10-01

    This report discusses the second phase of the aging assessment of nuclear plant service water systems (SWSs) which was performed by the Pacific Northwest Laboratory (PNL) to support the US Nuclear Regulatory Commission's (NRC's) Nuclear Plant Aging Research (NPAR) program. The SWS was selected for study because of its essential role in the mitigation of and recovery from accident scenarios involving the potential for core-melt, and because it is subject to a variety of aging mechanisms. The objectives of the SWS task under the NPAR program are to identify and characterize the principal age-related degradation mechanisms relevant to this system, to assess the impact of aging degradation on operational readiness, and to provide a methodology for the management of aging on the service water aspect of nuclear plant safety. The primary degradation mechanism in the SWSs as stated in the Phase I assessment and confirmed by the analysis in Phase II, is corrosion compounded by biologic and inorganic accumulation. It then follows that the most effective means for mitigating degradation in these systems is to pursue appropriate programs to effectively control the water chemistry properties when possible and to use biocidal agents where necessary. A methodology for producing a complete root-cause analysis was developed as a result of needs identified in the Phase I assessment for a more formal procedure that would lend itself to a generic, standardized approach. It is recommended that this, or a similar methodology, be required as a part of the documentation for corrective maintenance performed on the safety-related portions of SWSs to provide an accurate focus for effective management of aging

  18. Ageing degradation in the Gentilly-1 concrete containment building

    International Nuclear Information System (INIS)

    Jaffer, S.; Pentecost, S.; Angell, P.; Shenton, B.

    2015-01-01

    Concrete containment buildings (CCBs) are designed for a service life up to 40 years, but nuclear power plant (NPP) refurbishment can extend service life beyond 60 years. Only limited testing can be conducted on an in-service CCB. The Gentilly-1 (G-1) NPP is in a safe, sustainable shutdown state and the G-1 CCB was available for testing to determine age-related degradation that may be relevant to operating CCBs. Visual observation of the G-1 CCB helped to identify various signs of degradation. However, field testing, via concrete removal, was performed to: (i) examine reinforcing bars and concrete to determine their condition and in-situ stresses and (ii) examine condition of post-tensioned (P-T) wires. The concrete was also subjected to laboratory tests to evaluate its physical, mechanical and chemical properties such as compressive strength, carbonation depth, chloride content and presence of internal degradation. The degradation mechanisms that were clearly visible include macro- and micro-cracking, efflorescence, and weathering. The reinforcing bars in the perimeter wall and dome exposed during the program showed no evidence of active corrosion. Corrosion products were observed on the surfaces of most exposed P-T wires in the perimeter wall, but none were present on P-T wires exposed in the dome. Laboratory testing on the concrete cores extracted from the CCB revealed compressive strength in excess of the design requirements, low carbonation depths (< 10 mm) and no appreciable chlorides. Micro-cracking was observed in the samples recovered from the wall and dome. To date, the observed micro-cracking has had no apparent visible affect on the performance of the CCB concrete. (authors)

  19. Early detection of materials degradation

    Science.gov (United States)

    Meyendorf, Norbert

    2017-02-01

    Lightweight components for transportation and aerospace applications are designed for an estimated lifecycle, taking expected mechanical and environmental loads into account. The main reason for catastrophic failure of components within the expected lifecycle are material inhomogeneities, like pores and inclusions as origin for fatigue cracks, that have not been detected by NDE. However, material degradation by designed or unexpected loading conditions or environmental impacts can accelerate the crack initiation or growth. Conventional NDE methods are usually able to detect cracks that are formed at the end of the degradation process, but methods for early detection of fatigue, creep, and corrosion are still a matter of research. For conventional materials ultrasonic, electromagnetic, or thermographic methods have been demonstrated as promising. Other approaches are focused to surface damage by using optical methods or characterization of the residual surface stresses that can significantly affect the creation of fatigue cracks. For conventional metallic materials, material models for nucleation and propagation of damage have been successfully applied for several years. Material microstructure/property relations are well established and the effect of loading conditions on the component life can be simulated. For advanced materials, for example carbon matrix composites or ceramic matrix composites, the processes of nucleation and propagation of damage is still not fully understood. For these materials NDE methods can not only be used for the periodic inspections, but can significantly contribute to the material scientific knowledge to understand and model the behavior of composite materials.

  20. Environmental Degradation: Causes and Consequences

    Directory of Open Access Journals (Sweden)

    Swati Tyagi

    2014-08-01

    Full Text Available The subject of environmental economics is at the forefront of the green debate: the environment can no longer be viewed as an entity separate from the economy. Environmental degradation is of many types and have many consequences. To address this challenge a number of studies have been conducted in both developing and developed countries applying different methods to capture health benefits from improved environmental quality. Minimizing exposure to environmental risk factors by enhancing air quality and access to improved sources of drinking and bathing water, sanitation and clean energy is found to be associated with significant health benefits and can contribute significantly to the achievement of the Millennium Development Goals of environmental sustainability, health and development. In this paper, I describe the national and global causes and consequences of environmental degradation and social injustice. This paper provides a review of the literature on studies associated with reduced environmental risk and in particular focusing on reduced air pollution, enhanced water quality and climate change mitigation.