WorldWideScience

Sample records for in-vessel melt water

  1. Numerical modeling of in-vessel melt water interaction in large scale PWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Siemens AG, KWU NA-M, Erlangen (Germany)

    1998-01-01

    This paper presents a comparison between IVA4 simulations and FARO L14, L20 experiments. Both experiments were performed with the same geometry but under different initial pressures, 51 and 20 bar respectively. A pretest prediction for test L21 which is intended to be performed under an initial pressure of 5 bar is also presented. The strong effect of the volume expansion of the evaporating water at low pressure is demonstrated. An in-vessel simulation for a 1500 MW el. PWR is presented. The insight gained from this study is: that at no time are conditions for the feared large scale melt-water intermixing at low pressure in force, with this due to the limiting effect of the expansion process which accelerates the melt and the water into all available flow paths. (author)

  2. In vessel core melt progression phenomena

    International Nuclear Information System (INIS)

    Courtaud, M.

    1993-01-01

    For all light water reactor (LWR) accidents, including the so called severe accidents where core melt down can occur, it is necessary to determine the amount and characteristics of fission products released to the environment. For existing reactors this knowledge is used to evaluate the consequences and eventual emergency plans. But for future reactors safety authorities demand decrease risks and reactors designed in such a way that fission products are retained inside the containment, the last protective barrier. This requires improved understanding and knowledge of all accident sequences. In particular it is necessary to be able to describe the very complex phenomena occurring during in vessel core melt progression because they will determine the thermal and mechanical loads on the primary circuit and the timing of its rupture as well as the fission product source term. On the other hand, in case of vessel failure, knowledge of the physical and chemical state of the core melt will provide the initial conditions for analysis of ex-vessel core melt progression and phenomena threatening the containment. Finally a good understanding of in vessel phenomena will help to improve accident management procedures like Emergency Core Cooling System water injection, blowdown and flooding of the vessel well, with their possible adverse effects. Research and Development work on this subject was initiated a long time ago and is still in progress but now it must be intensified in order to meet the safety requirements of the next generation of reactors. Experiments, limited in scale, analysis of the TMI 2 accident which is a unique source of global information and engineering judgment are used to establish and assess physical models that can be implemented in computer codes for reactor accident analysis

  3. Oxidation effects during corium melt in-vessel retention

    Energy Technology Data Exchange (ETDEWEB)

    Almyashev, V.I.; Granovsky, V.S.; Khabensky, V.B.; Krushinov, E.V.; Sulatsky, A.A.; Vitol, S.A. [Alexandrov Scientific-Research Institute of Technology (NITI), Sosnovy Bor (Russian Federation); Gusarov, V.V. [Ioffe Institute, St. Petersburg (Russian Federation); Bechta, S. [Royal Institute of Technology (KHT), Stockholm (Sweden); Barrachin, M.; Fichot, F. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), St Paul lez Durance (France); Bottomley, P.D., E-mail: paul.bottomley@ec.europa.eu [Joint Research Centre, Institut für Transurane (ITU), Karlsruhe (Germany); Fischer, M. [AREVA GmbH, Erlangen (Germany); Piluso, P. [CEA Cadarache-DEN/DTN/STRI (France)

    2016-08-15

    Highlights: • Corium–steel interaction tests were re-examined particularly for transient processes. • Oxidation of corium melt was sensitive to oxidant supply and surface characteristics. • Consequences for vessel steel corrosion rates in severe accidents were discussed. - Abstract: In the in-vessel corium retention studies conducted on the Rasplav-3 test facility within the ISTC METCOR-P project and OECD MASCA program, experiments were made to investigate transient processes taking place during the oxidation of prototypic molten corium. Qualitative and quantitative data have been produced on the sensitivity of melt oxidation rate to the type of oxidant, melt composition, molten pool surface characteristics. The oxidation rate is a governing factor for additional heat generation and hydrogen release; also for the time of secondary inversion of oxidic and metallic layers of corium molten pool.

  4. The modeling of core melting and in-vessel corium relocation in the APRIL code

    Energy Technology Data Exchange (ETDEWEB)

    Kim. S.W.; Podowski, M.Z.; Lahey, R.T. [Rensselaer Polytechnic Institute, Troy, NY (United States)] [and others

    1995-09-01

    This paper is concerned with the modeling of severe accident phenomena in boiling water reactors (BWR). New models of core melting and in-vessel corium debris relocation are presented, developed for implementation in the APRIL computer code. The results of model testing and validations are given, including comparisons against available experimental data and parametric/sensitivity studies. Also, the application of these models, as parts of the APRIL code, is presented to simulate accident progression in a typical BWR reactor.

  5. Study on severe fuel damage and in-vessel melt progression

    International Nuclear Information System (INIS)

    Kim, Hee Dong; Kim, Sang Baik; Lee, Gyu Jung

    1992-06-01

    In-vessel core melt progression describes the progression of the state of a reactor core from core uncovery up to reactor vessel melt through in uncovered accidents or through temperature stabilization in accidents recovered by core reflooding. Melt progression can be thought as two parts; early melt progression and late melt progression. Early phase of core melt progression includes the progression of core material melting and relocation, which mostly consist of metallic materials. On the other hand, the late phase of core melt progression involves ceramic material melt and relocation to the lower plenum and heat-up the reactor vessel lower head. A large number of information are available for the early melt progression through experiments such as SFD, DF, FLHT test and utilized in the severe accident analysis codes. However, understanding of the late phase melt progression phenomenology is based primary on TMI-2 core examinations and not much experimental information is available. Especilally, the great uncertainties exist in vessel failure mode, melt composition, mass, and temperature. Further research is planned to perform to reduce the uncertainties in understanding of core melt down accidents as parts of long term melt progression research program. A study on the core melt progression at KAERI has been being performed through the Severe Accident Research Program with USNRC. KAERI staff had participated in the PBF SFD experiments at INEL and analyses of experiments were performed using SCDAP code. Experiments of core melt program have not been carried out at KAERI yet. It is planned that further research on core melt down accidents will be performed, which is related to design of future generations of nuclear reactors as parts of long-term project for improvement of nuclear reactor safety. (Author)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

    Full text of publication follows: LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with uniform gap of 5 mm or 10 mm to the inner surface of the lower head vessel. As a performance test of the in-vessel core catcher, the effects of base steel and internal coating materials and gap thickness between the core catcher and the lower head vessel were examined in this study. In the LAVA-GAP-2 and LAVA-GAP-3 tests, the base steel was carbon steel and the gap thickness was 10 mm. On the other hand, in the LAVA-GAP-4 and LAVA-GAP-5 tests, the base steel was stainless steel and the gap thickness was 5 mm. Actual composition of the coating material for the LAVA-GAP-4 test was 92% of ZrO 2 - 8% of Y 2 O 3 including 95% of Ni - 5% of Al bond coat same as the LAVA-GAP-3 test. In these tests, the thickness of ZrO 2 internal coating was 0.5 mm. To examine the effects of the coating material, in-vessel core catcher with a 0.6 mm-thick ZrO 2 coating without bond coat was used in the LAVA-GAP-5 test. This report summarizes the experimental results and the post metallurgical inspection results of the LAVA-GAP-4 and LAVA-GAP- 5 tests. In the LAVA-GAP-4 and LAVA-GAP-5 tests, the core catcher was failed and it was stuck to the inner surface of the lower head vessel. LAVA-GAP-4 and LAVA-GAP-5 test results imply that 5 mm thick gap is rather small for sufficient water ingression and steam venting through the gap. In case of small gap size, water is boiled off and steam increases pressure inside the gap and so water can not ingress into the gap at the initial heat up stage. Metallurgical inspections on the test specimens indicate that the internal coating layer might melt totally and dispersed in the base steel and the solidified iron melt and so the detection frequencies of Zr and O are trivial all

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    2009-04-01

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

  10. SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kolev, N.I. [Framatome Advanced Nuclear Power, NDSI, Erlangen (Germany)

    2001-07-01

    Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

  11. SWR 1000 severe accident control through in-vessel melt retention by external RPV cooling

    International Nuclear Information System (INIS)

    Kolev, N.I.

    2001-01-01

    Framatome Advanced Nuclear Power is being designing a new generation NPP with boiling water reactor SWR1000. Besides of various of modern passive and active safety features the system is also designed for controlling of a postulated severe accident with extreme low probability of occurrence. This work presents the rationales behind the decision to select the external cooling as a safety management strategy during severe accident. Bounding scenery are analyzed regarding the core melting, melt-water interaction during relocation of the melt from the core region into the lower head and the external coolability of the lower head. The conclusion is reached that the external cooling for the SWR1000 is a valuable strategy for accident management during postulated severe accidents. (authors)

  12. Analysis of In-Vessel Late Phase Melt Progression Using SCDAP/RELAP5/MOD3.3

    International Nuclear Information System (INIS)

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

    2004-01-01

    High-pressure in-vessel melt progressions of the KSNP (Korean Standard Nuclear Power Plant) have been analyzed using the SCDAP/RELAP5/MOD3.3 computer code. The total loss of feed water (LOFW) to the steam generators with/without intentional RCS depressurization using the safety depressurization system (SDS) and the station blackout (SBO) have been simulated from transient initiation to reactor vessel failure. The SCDAP/RELAP5/MOD3.3 results have shown that the pressure boundary of the reactor coolant system did not fail before reactor vessel failure in the high-pressure sequences of the LOFW and the SBO transients of the KSNP. In all the high-pressure transients, approximately 20-30 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure. Intentional RCS depressurization using the SDS for the total LOFW delays reactor vessel failure for approximately 5 hours by actuation of the safety injection tanks. At the time of reactor vessel failure, approximately 50-60 % of the fuel rod cladding was oxidized for the total LOFW and the SBO transients of the KSNP. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  14. Studies on melt-water-structure interaction during severe accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.; Okkonen, T.J.; Bui, V.A.; Nourgaliev, R.R.; Andersson, J.

    1996-10-01

    Results of a series of studies, on melt-water-structure interactions which occur during the progression of a core melt-down accident, are described. The emphasis is on the in-vessel interactions and the studies are both experimental and analytical. Since, the studies performed resulted in papers published in proceedings of the technical meetings, and in journals, copies of a set of selected papers are attached to provide details. A summary of the results obtained is provided for the reader who does not, or cannot, venture into the perusal of the attached papers. (au)

  15. Studies on melt-water-structure interaction during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Dinh, T.N.; Okkonen, T.J.; Bui, V.A.; Nourgaliev, R.R.; Andersson, J. [Royal Inst. of Technology, Div. of Nucl. Power Safety, Stockholm (Sweden)

    1996-10-01

    Results of a series of studies, on melt-water-structure interactions which occur during the progression of a core melt-down accident, are described. The emphasis is on the in-vessel interactions and the studies are both experimental and analytical. Since, the studies performed resulted in papers published in proceedings of the technical meetings, and in journals, copies of a set of selected papers are attached to provide details. A summary of the results obtained is provided for the reader who does not, or cannot, venture into the perusal of the attached papers. (au).

  16. Water jet intrusion into hot melt concomitant with direct-contact boiling of water

    Energy Technology Data Exchange (ETDEWEB)

    Sibamoto, Yasuteru [Japan Atomic Energy Research Inst., Tokai Research Establishment, Tokai, Ibaraki (Japan)

    2005-08-01

    Boiling of water poured on surface of high-temperature melt (molten metal or metal oxide) provides an efficient means for heat exchange or cooling of melt. The heat transfer surface area can be extended by forcing water into melt. Objectives of the present study are to elucidate key factors of the thermal and hydrodynamic interactions for the water jet injection into melt (Coolant Injection mode). Proposed applications include in in-vessel heat exchangers for liquid metal reactor and emergency measures for cooling of molten core debris in severe accidents of light water reactor. Water penetration into melt may occurs also as a result of fuel-coolant interaction (FCI) in modes other than CI, it is anticipated that the present study contributes to understand the fundamental mechanism of the FCI process. The previous works have been limited on understanding the melt-water interaction phenomena in the water-injection mode because of difficulty in experimental measurement where boiling occurs in opaque invisible hot melt unlike the melt-injection mode. We conducted visualization and measurement of melt-water-vapor multiphase flow phenomena by using a high-frame-rate neutron radiography technique and newly-developed probes. Although limited knowledge, however, has been gained even such an approach, the experimental data were analyzed deeply by comparing with the knowledge obtained from relevant matters. As a result, we succeeded in revealing several key phenomena and validity in the conditions under which stable heat transfer is established. Moreover, a non-intrusive technique for measurement of the velocity and pressure fields adjacent to a moving free surface is developed. The technique is based on the measurement of fluid surface profile, which is useful for elucidation of flow mechanism accompanied by a free surface like the present phenomena. (author)

  17. Validation of the THIRMAL-1 melt-water interaction code

    Energy Technology Data Exchange (ETDEWEB)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W. [Argonne National Lab., IL (United States)

    1995-09-01

    The THIRMAL-1 computer code has been used to calculate nonexplosive LWR melt-water interactions both in-vessel and ex-vessel. To support the application of the code and enhance its acceptability, THIRMAL-1 has been compared with available data from two of the ongoing FARO experiments at Ispra and two of the Corium Coolant Mixing (CCM) experiments performed at Argonne. THIRMAL-1 calculations for the FARO Scoping Test and Quenching Test 2 as well as the CCM-5 and -6 experiments were found to be in excellent agreement with the experiment results. This lends confidence to the modeling that has been incorporated in the code describing melt stream breakup due to the growth of both Kelvin-Helmholtz and large wave instabilities, the sizes of droplets formed, multiphase flow and heat transfer in the mixing zone surrounding and below the melt metallic phase. As part of the analysis of the FARO tests, a mechanistic model was developed to calculate the prefragmentation as it may have occurred when melt relocated from the release vessel to the water surface and the model was compared with the relevant data from FARO.

  18. 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.

  19. 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

  20. 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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  2. Experiments and analyses on melt-structure-water interactions during severe accidents

    International Nuclear Information System (INIS)

    Seghal, B.R.; Dinh, T.N.; Bui, V.A.; Green, J.A.; Nourgaliev, R.R.; Okkonen, T.O.; Dinh, A.T.

    1998-04-01

    This report is the final report for the research project Melt Structure Water Interactions (MSWI). It describes results of analytical and experimental studies concerning MSWI during the course of a hypothetical core meltdown accident in a LWR. Emphasis has been placed on phenomena which govern vessel failure mode and timing and the mechanisms and properties which govern the fragmentation and breakup of melt jets and droplets. It was found that: 2-D effects significantly diminished the focusing effect of an overlying metallic layer on top of an oxide melt pool. This result improves the feasibility of in-vessel retention of a melt pool through external cooling of the lower head; phenomena related to hole ablation and melt discharge, in the event of vessel failure, are affected significantly by crust formation; the jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters but also on the melt physical properties, which change as the melt cools down from liquid to solid temperature; film boiling was investigated by developing a two-phase flow model and inserting it in a multi-D fluid dynamics code. It was concluded that the thickness of the film on the surface of a melt jet would be small and that the effects of the film on the process should not be large. This conclusion is contrary to the modeling employed in some other codes. The computer codes were developed and validated against the data obtained in the MSWI Project. The melt vessel interaction thermal analysis code describes the process of melt pool formation and convection and the resulting vessel thermal loadings. In addition, several innovative models were developed to describe the melt-water interaction process. The code MELT-3D treats the melt jet as a collection of particles whose movement is described with a three-dimensional Eulerian formulation. The model (SIPHRA) tracks the melt jet with an additional equation, using the

  3. Reaction of soda-lime-silica glass melt with water vapour at melting temperatures

    Czech Academy of Sciences Publication Activity Database

    Vernerová, Miroslava; Kloužek, Jaroslav; Němec, Lubomír

    2015-01-01

    Roč. 416, MAY 15 (2015), s. 21-30 ISSN 0022-3093 R&D Projects: GA TA ČR TA01010844 Institutional support: RVO:67985891 Keywords : glass melt * sulfate * water vapour * bubble nucleation * melt foaming * glass melting Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 1.825, year: 2015

  4. Fragmentation and quench behavior of corium melt streams in water

    International Nuclear Information System (INIS)

    Spencer, B.W.; Wang, K.; Blomquist, C.A.; McUmber, L.M.; Schneider, J.P.

    1994-02-01

    The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (i) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (ii) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (iii) the quench rate of the molten fuel through the water in the lower plenum, (iv) the steam generation and hydrogen generation during the interaction, (v) the transient pressurization of the primary system, and (vi) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics

  5. Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Park, H.S.; Giri, A.; Karbojian, A.; Jasiulevicius, A.; Hansson, R.C.; Chikkanagoudar, U.; Shiferaw, D.; Stepanyan, A.

    2004-01-01

    This is the annual report for the work performed in year 2003 in the research project 'Melt-Structure-Water Interactions (MSWI) During Severe Accidents in LWRs', under the auspices of the APRI Project, jointly funded by SKI, HSK, and the Swedish and Finnish power companies. The emphasis of the work was placed on phenomena and parameters, which govern the droplet fragmentation in steam explosions, in-vessel and ex-vessel melt/debris coolability, melt pool convection, and the thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Most research projects in 2002, such as the COMECO, POMECO and MISTEE programs, were continued. An analysis of the FOREVER experiments using the RELAP code to investigate the melt coolability, bubble dynamics and bubble stability to investigate the dynamic behavior of vapor bubble during steam explosions and associated melt fragmentation, quenching boiling experiment to investigate the thermal behavior of single melt droplet were newly initiated. The SIMECO experiment to investigate the three-layer melt pool convection was restarted. The experimental facilities for these projects were fully functional during year 2003. Many of the investigations performed during the course of the MSWI project have produced papers, which have been published in the proceedings of technical meetings and Journals. Significant technical advances were achieved during the course of these studies. These were: A series of experiments on single drop steam explosions was performed to investigate the fine fragmentation process of a metallic melt drop in various thermal conditions. For the first time, transient fine fragmentation process of a melt drop during explosion phase of a steam explosion was visualized continuously and quantified. Different triggering behavior with respect to the coolant subcooling was observed. The analyses on bubble dynamics during a single drop steam explosion and vapor bubble stability estimated the dynamic

  6. Cloud screening and melt water detection over melting sea ice using AATSR/SLSTR

    Science.gov (United States)

    Istomina, Larysa; Heygster, Georg

    2014-05-01

    With the onset of melt in the Arctic Ocean, the fraction of melt water on sea ice, the melt pond fraction, increases. The consequences are: the reduced albedo of sea ice, increased transmittance of sea ice and affected heat balance of the system with more heat passing through the ice into the ocean, which facilitates further melting. The onset of melt, duration of melt season and melt pond fraction are good indicators of the climate state of the Arctic and its change. In the absence of reliable sea ice thickness retrievals in summer, melt pond fraction retrieval from satellite is in demand as input for GCM as an indicator of melt state of the sea ice. The retrieval of melt pond fraction with a moderate resolution radiometer as AATSR is, however, a non-trivial task due to a variety of subpixel surface types with very different optical properties, which give non-unique combinations if mixed. In this work this has been solved by employing additional information on the surface and air temperature of the pixel. In the current work, a concept of melt pond detection on sea ice is presented. The basis of the retrieval is the sensitivity of AATSR reflectance channels 550nm and 860nm to the amount of melt water on sea ice. The retrieval features extensive usage of a database of in situ surface albedo spectra. A tree of decisions is employed to select the feasible family of in situ spectra for the retrieval, depending on the melt stage of the surface. Reanalysis air temperature at the surface and brightness temperature measured by the satellite sensor are analyzed in order to evaluate the melting status of the surface. Case studies for FYI and MYI show plausible retrieved melt pond fractions, characteristic for both of the ice types. The developed retrieval can be used to process the historical AATSR (2002-2012) dataset, as well as for the SLSTR sensor onboard the future Sentinel-3 mission (scheduled for launch in 2015), to keep the continuity and obtain longer time sequence

  7. Water-fluxed melting of the continental crust: A review

    Czech Academy of Sciences Publication Activity Database

    Weinberg, R. F.; Hasalová, Pavlína

    212-215, January (2015), s. 158-188 ISSN 0024-4937 Institutional support: RVO:67985530 Keywords : aqueous fluids * crustal anatexis * granites * silicate melts * water-fluxed melting Subject RIV: DB - Geology ; Mineralogy Impact factor: 3.723, year: 2015

  8. Experiments on melt droplets falling into a water pool

    Energy Technology Data Exchange (ETDEWEB)

    Okkonen, T.; Sehgal, B.R. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1998-01-01

    This paper presents experimental data and analysis related to melt droplets falling into a water pool. A binary CaO-B{sub 2}O{sub 3} melt mixture is used to study the influence of melt superheat and water subcooling on droplet deformation and fragmentation. For the conditions studied (We {<=} 1000), the surface tension of the melt droplet and the film boiling stability greatly affect the fragmentation behaviour. If the melt temperature is between the liquidus and solidus point (mushy zone) or if the film boiling is stable due to a relatively low subcooling, the droplet deformation and fragmentation are mitigated. This behaviour can be related to the effective Weber number (We) of the melt droplet upon entry into the water pool. Similar phenomena can be expected also for interactions of corium (UO{sub 2}-ZrO{sub 2}) and water, which are characterized by a potentially fast transformation of melt into the mushy zone and by particularly stable film boiling. (author)

  9. Heat dissipation research on the water-cooling channel of HL-2M in-vessel coils

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J., E-mail: jiangjiaming@swip.ac.cn; Liu, Y.; Chen, Q.; Ji, X.Q.

    2017-04-15

    Highlights: • The joule heat of in-vessel coils is very difficult to dissipate inside HL-2M vacuum vessel. • Heat dissipation model of the coil includes the joule heat model, the heat conduction model and the heat transfer model. • The CFD analysis has been done for the coil-water cooling, with comparison with the date of theoretical analysis and experiment. • The result shows water-cooling channel is good for the joule heat transfer and taken away. - Abstract: HL-2M in-vessel coils are positioned in high vacuum circumstance, and they will generate joule heat when they carry 15 kA electrical current, but joule heat is very difficult to dissipate in vacuum, so a hollow cable with 8 mm inner diameter is design as water-cooling channel for heat convection. By using the methods of the theoretical derivation, together with CFD numeric simulation method and the experiment of the heat transfer, the water channel of HL-2M in-vessel coils has been studied, and the temperature of HL-2M in-vessel coils under different cooling water flow rates is obtained and acceptable. Simultaneously, the external cooling water supply system parameters for the water-cooling channel of the coils are estimated. Three methods’ results are in good agreement; the theoretical model is verified and could be popularized for predicting the temperature rise of HL-2M in-vessel coils.

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

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

  12. Direct contact heat transfer characteristics between melting alloy and water

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Nishi, Yoshihisa; Furuya, Masahiro

    1995-01-01

    As a candidate for an innovative steam generator for fast breeder reactors, a heat exchanger with direct contact heat transfer between melting alloy and water was proposed. The evaluation of heat transfer characteristics of this heat exchanger is one of the research subjects for the design and development of the steam generator. In this study, the effect of the pressure on heat transfer characteristics and the required degree of superheating of melting alloy above water saturation temperature are evaluated during the direct contact heat transfer experiment by injecting water into Wood's alloy. In the experiment, the pressure, the temperature of the Wood's alloy, the flow rate of feed water, and the depth of the feed water injection point are varied as parameters. As a result of the experiment, the product of the degree of Wood's alloy superheating above water saturation temperature and the depth of the feed water injection point is constant for each pressure. This constant increases as the pressure rises. (author)

  13. Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

    Directory of Open Access Journals (Sweden)

    J. Mao

    2018-03-01

    Full Text Available The strategy denoted as in-vessel retention (IVR is widely used in severe accident (SA management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

  14. Developing a Hygrometer for Water-Undersaturated Lherzolite Melts

    Science.gov (United States)

    Guild, M. R.; Till, C. B.

    2017-12-01

    The effect of water on the composition of primitive mantle melts at arc volcanoes is a topic of wide interest and has been addressed in a number of previous experimental studies including Hirose & Kawamoto (1995), Gaetani & Grove (1998), Till et al. (2012) and Mitchell & Grove (2015). The current study builds upon the work by previous authors in an effort to develop a more robust hygrometer for primitive lherzolite melts at water-undersaturated conditions. The starting composition for this experimental study is a mixture of 75% primitive upper mantle and 25% primitive basalt (Baker et al., 1991) with a bulk H2O content of 2 wt. %. Experiments were performed at Arizona State University in the Experimental Petrology and Igneous processes Center (EPIC) from 1.2-1.6 GPa at 1150-1300 ºC for 2 days in a piston cylinder apparatus to reflect conditions relevant for arc melt equilibration (Till 2017). A double capsule design was used to prevent Fe and H2O loss with an inner Fe-presaturated Au80Pd20 capsule and an outer Au80Pd20 capsule. Run products were analyzed by electron microprobe and determined to be successful when they demonstrated 0-5% Fe-loss, olivine-melt KDs of 0.27-0.30, and minimal H2O loss. The water-undersaturated melt composition are in equilibrium with ol+opx+sp±cpx. Run products at 1.6 GPa do not contain cpx in the mineral assemblage over the studied temperature range. Observed melt compositions have SiO2 contents of 48-49 wt. % at 1.2 GPa and 46-49 wt.% at 1.6 GPa. Our experimental results suggest an enhanced effect of water on increasing the SiO2 content of the melt compared to previous studies on systems with similar water contents and anhydrous systems. Baker, et al., JGR 96, 21819-21842 (1991). Gaetani & Grove, CMP 131, 323-346 (1998). Hirose & Kawamoto, EPSL 133, 463-473 (1995). Mitchell & Grove, CMP 170, 13 (2015). Till, Am. Mineral, 102, 931-947 (2017). Till, et al., JGR 117 (2012).

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

    International Nuclear Information System (INIS)

    Kolev, Nikolay Ivanov

    2004-01-01

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

  16. Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

    International Nuclear Information System (INIS)

    Additon, S.L.; Fontana, M.H.; Carter, J.C.

    1990-01-01

    This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure

  17. Snow cover as a source of technogenic pollution of surface water during the snow melting period

    OpenAIRE

    Labuzova Olga; Noskova Tatyana; Lysenko Maria; Ovcharenko Elena; Papina Tatyana

    2016-01-01

    The study of pollutants in melt water of snow cover and snow disposal sites in the city of Barnaul showed that during the snow melting period the surface water is not subjected to significant technogenic impact according to a number of studied indices. The oils content is an exception: it can exceed MAC more than 20 times in river- water due to the melting of city disposal sites. Environmental damage due to an oils input into water resources during the snow melting period...

  18. Analysis of Water Recovery Rate from the Heat Melt Compactor

    Science.gov (United States)

    Balasubramaniam, R.; Hegde, U.; Gokoglu, S.

    2013-01-01

    Human space missions generate trash with a substantial amount of plastic (20% or greater by mass). The trash also contains water trapped in food residue and paper products and other trash items. The Heat Melt Compactor (HMC) under development by NASA Ames Research Center (ARC) compresses the waste, dries it to recover water and melts the plastic to encapsulate the compressed trash. The resulting waste disk or puck represents an approximately ten-fold reduction in the volume of the initial trash loaded into the HMC. In the current design concept being pursued, the trash is compressed by a piston after it is loaded into the trash chamber. The piston face, the side walls of the waste processing chamber and the end surface in contact with the waste can be heated to evaporate the water and to melt the plastic. Water is recovered by the HMC in two phases. The first is a pre-process compaction without heat or with the heaters initially turned on but before the waste heats up. Tests have shown that during this step some liquid water may be expelled from the chamber. This water is believed to be free water (i.e., not bound with or absorbed in other waste constituents) that is present in the trash. This phase is herein termed Phase A of the water recovery process. During HMC operations, it is desired that liquid water recovery in Phase A be eliminated or minimized so that water-vapor processing equipment (e.g., condensers) downstream of the HMC are not fouled by liquid water and its constituents (i.e., suspended or dissolved matter) exiting the HMC. The primary water recovery process takes place next where the trash is further compacted while the heated surfaces reach their set temperatures for this step. This step will be referred to herein as Phase B of the water recovery process. During this step the waste chamber may be exposed to different selected pressures such as ambient, low pressure (e.g., 0.2 atm), or vacuum. The objective for this step is to remove both bound and

  19. Forced convective melting at an evolving ice-water interface

    Science.gov (United States)

    Ramudu, Eshwan; Hirsh, Benjamin; Olson, Peter; Gnanadesikan, Anand

    2015-11-01

    The intrusion of warm Circumpolar Deep Water into the ocean cavity between the base of ice shelves and the sea bed in Antarctica causes melting at the ice shelves' basal surface, producing a turbulent melt plume. We conduct a series of laboratory experiments to investigate how the presence of forced convection (turbulent mixing) changes the delivery of heat to the ice-water interface. We also develop a theoretical model for the heat balance of the system that can be used to predict the change in ice thickness with time. In cases of turbulent mixing, the heat balance includes a term for turbulent heat transfer that depends on the friction velocity and an empirical coefficient. We obtain a new value for this coefficient by comparing the modeled ice thickness against measurements from a set of nine experiments covering one order of magnitude of Reynolds numbers. Our results are consistent with the altimetry-inferred melting rate under Antarctic ice shelves and can be used in climate models to predict their disintegration. This work was supported by NSF grant EAR-110371.

  20. Experimental study of the fragmentation and quench behavior of corium melts in water

    International Nuclear Information System (INIS)

    Wang, S.K.; Blomquist, C.A.; Spencer, B.W.; McUmber, L.M.; Schneider, J.P.; Illinois Univ., Urbana, IL

    1989-01-01

    The interaction of molten core materials with water has been investigated for the pour stream mixing mode. This interaction plays a crucial role during the later stages of in-vessel core melt progression inside a light water reactor such as during the TMI-2 accident. The key issues which arise during the molten core relocation include: (1) the thermal attack and possible damage to the RPV lower head from the impinging molten fuel stream and/or the debris bed, (2) the molten fuel relocation pathways including the effects of redistribution due to core support structure and the reactor lower internals, (3) the quench rate of the molten fuel through the water in the lower plasma, (4) the steam generation and hydrogen generation during the interaction, (5) the transient pressurization of the primary system, and (6) the possibility of a steam explosion. In order to understand these issues, a series of six experiments (designated CCM-1 through -6) was performed in which molten corium passed through a deep pool of water in a long, slender pour stream mode. Results discussed include the transient temperatures and pressures, the rate and magnitude of steam/hydrogen generation, and the posttest debris characteristics. 9 refs., 29 figs

  1. Water boiling on the corium melt surface under VVER severe accident conditions

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B.; Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y.

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO 2+x -16% ZrO 2 -15% Fe 2 O 3 -6% Cr 2 O 3 -3% Ni 2 O 3 . The melt surface temperature ranged within 1920-1970 K. (orig.)

  2. Water boiling on the corium melt surface under VVER severe accident conditions

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.

    1999-01-01

    Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO 2 - 16%ZrO 2 - 15%Fe 2 O 3 - 6%Cr 2 O 3 -3%Ni 2 O 3 . The melt surface temperature was 1650-1700degC. (author)

  3. Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents: Year 2004 Activity. APRI 5 report

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-01

    This report presents descriptions of the major results obtained in the research program 'Melt-Structure-Water Interaction (MSWI)' at NPS/RIT during the year 2004. The primary objectives of the MSWI Project in year 2004 were to study (1) the in-vessel and exvessel melt/debris bed coolability process when melt is flooded with water, and (2) the energetics and characteristics of steam explosions. Our general approaches are to establish scaling relationships so that the data obtained in the experiments could be extended to prototypical accident geometries and conditions, develop phenomenological or computational models for the processes under investigation and validate the existing and newly developed models against data obtained at RIT and at other laboratories. In 2004, several experimental programs, such as the COMECO (Corium MElt COolability), POMECO (POrous MEdia COolability) and MISTEE (Micro-Interactions in STeam Explosion Experiments) programs were continued. The SIMECO (Simulation of MElt Coolability) program was restarted in 2004. The construction of the POMECO-GRAND (POrous MEdia COolability) facility was delayed due to lack of finances. However, existing POMECO facility was modified to study 3-D effects on debris coolability. In this report, the results from the COMECO experiment with high temperature oxidic melt, from the POMECO experiments for the multi-dimensional effects on debris bed coolability, from the SIMECO experiment for three-layer pool configuration and from the MISTEE experiments for steam explosion characteristics and loads are described. For analytical efforts, results from the COMETA code for the entire process of the steam explosions are discussed.

  4. Snow cover as a source of technogenic pollution of surface water during the snow melting period

    Directory of Open Access Journals (Sweden)

    Labuzova Olga

    2016-10-01

    Full Text Available The study of pollutants in melt water of snow cover and snow disposal sites in the city of Barnaul showed that during the snow melting period the surface water is not subjected to significant technogenic impact according to a number of studied indices. The oils content is an exception: it can exceed MAC more than 20 times in river- water due to the melting of city disposal sites. Environmental damage due to an oils input into water resources during the snow melting period can be more than 300000 thousand rubles.

  5. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    International Nuclear Information System (INIS)

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path

  6. Effect of boiling regime on melt stream breakup in water

    International Nuclear Information System (INIS)

    Spencer, B.W.; Gabor, J.D.; Cassulo, J.C.

    1986-01-01

    A study has been performed examining the breakup and mixing behavior of an initially coherent stream of high-density melt as it flows downward through water. This work has application to the quenching of molten core materials as they drain downward during a postulated severe reactor accident. The study has included examination of various models of breakup distances based upon interfacial instabilities dominated either by liquid-liquid contact or by liquid-vapor contact. A series of experiments was performed to provide a data base for assessment of the various modeling approaches. The experiments involved Wood's metal (T/sub m/ = 73 0 C, rho = 9.2 g/cm 3 , d/sub j/ = 20 mm) poured into a deep pool of water. The temperature of the water and wood's metal were varied to span the range from single-phase, liquid-liquid contact to the film boiling regime. Experiment results showed that breakup occurred largely as a result of the spreading and entrainment from the leading edge of the jet. However, for streams of sufficient lengths a breakup length could be discerned at which there was no longer a coherent central core of the jet to feed the leading edge region. The erosion of the vertical trailing column is by Kelvin-Helmoltz instabilities and related disengagement of droplets from the jet into the surrounding fluid. For conditions of liquid-liquid contact, the breakup length has been found to be about 20 jet diameters; when substantial vapor is produced at the interface due to heat transfer from the jet to the water, the breakup distance was found to range to as high as 50 jet diameters. The former values are close to the analytical prediction of Taylor, whereas the latter values are better predicted by the model of Epstein and Fauske

  7. Water boiling on the corium melt surface under VVER severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B. [Sci. Res. Technol. Inst., Leningrad (Russian Federation); Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y. [St. Petersburg Electrotechnical University (SPbEU), Prof. Popov st 5/3, St. Petersburg (Russian Federation)

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO{sub 2+x}-16% ZrO{sub 2}-15% Fe{sub 2}O{sub 3}-6% Cr{sub 2}O{sub 3}-3% Ni{sub 2}O{sub 3}. The melt surface temperature ranged within 1920-1970 K. (orig.)

  8. Water boiling on the corium melt surface under VVER severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V. [Research Institute of Technology, Sosnovy Bor (NITI) (RU)] [and others

    1999-07-01

    Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO{sub 2}- 16%ZrO{sub 2}- 15%Fe{sub 2}O{sub 3} - 6%Cr{sub 2}O{sub 3}-3%Ni{sub 2}O{sub 3}. The melt surface temperature was 1650-1700degC. (author)

  9. Water Recovery with the Heat Melt Compactor in a Microgravity Environment

    Science.gov (United States)

    Golliher, Eric L.; Goo, Jonathan; Fisher, John

    2015-01-01

    The Heat Melt Compactor is a proposed utility that will compact astronaut trash, extract the water for eventual re-use, and form dry square tiles that can be used as additional ionizing radiation shields for future human deep space missions. The Heat Melt Compactor has been under development by a consortium of NASA centers. The downstream portion of the device is planned to recover a small amount of water while in a microgravity environment. Drop tower low gravity testing was performed to assess the effect of small particles on a capillary-based water/air separation device proposed for the water recovery portion of the Heat Melt Compactor.

  10. A phenomenological analysis of melt progression in the lower head of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M., E-mail: jean-marie.seiler@cea.fr [CEA, DEN, DTN, F-38054 Grenoble (France); Tourniaire, B. [EDF/Septen, Lyon (France)

    2014-03-15

    Highlights: • We propose a phenomenological description of melt progression into the lower head. • We examine changes in heat loads on the vessel. • Heat loads are more severe than emphasized by the bounding situation assumption. • Both primary circuit and ex-vessel reflooding are necessary for in-vessel retention. • Vessel failure conditions are examined. - Abstract: The analysis of in-vessel corium cooling (IVC) and retention (IVR) involves the description of very complex and transient physical phenomena. To get round this difficulty, “bounding” situations are often emphasized for the demonstration of corium coolability, by vessel flooding and/or by reactor pit flooding. This approach however comes up against its own limitations. More realistic melt progression scenarios are required to provide plausible corium configurations and vessel failure conditions. Work to develop more realistic melt progression scenarios has been done at CEA, in collaboration with EDF. Development has concentrated on the French 1300 MWe PWR, considering both dry scenarios and the possibility of flooding of the RPC (reactor primary circuit) and/or the reactor pit. The models used for this approach have been derived from the analysis of the TMI2 accident and take benefit from the lessons derived from several programs related to pool thermal hydraulics (BALI, COPO, ACOPO, etc.), material interactions (RASPLAV, MASCA), critical heat flux (CHF) on the external surface of the vessel (KAIST, SULTAN, ULPU), etc. Important conclusions of this work are as follows: (a)After the start of corium melting and onset of melt formation in the core at low pressure (∼1 to 5 bars), it seems questionable that RPV (reactor pressure vessel) reflooding alone would be sufficient to achieve corium retention in the vessel; (b)If the vessel is not cooled externally, it may fail due to local heat-up before the whole core fuel inventory is relocated in the lower head; (c)Even if the vessel is

  11. TMI-2 in-vessel hydraulic systems utilize high water and high boron content fluids

    International Nuclear Information System (INIS)

    Baston, V.F.; Hofstetter, K.J.; Hofman, L.A.; Gallagher, R.E.

    1987-01-01

    Choice of a hydraulic fluid for use in the Three Mile Island Unit 2 (TMI-2) reactor vessel defueling equipment required consideration of the following constraints for the hydraulic fluid given an accidental spill into the reactor coolant system (RCS). The TMI-2 RCS hydraulic fluid utilized in the hydraulic operations utilized a solution composition of 95 wt% water and 5 wt% of the above base fluid. The TMI-2 hydraulic system utilizes pressures up to 3500 psi. The selected hydraulic fluid has been in use since December 1986 with minimal operational difficulties

  12. Melting of short 1-alcohol monolayers on water: Thermodynamics and x-ray scattering studies

    DEFF Research Database (Denmark)

    Berge, B.; Konovalov, O.; Lajzerowicz, J.

    1994-01-01

    From surface tension measurements we extract the melting entropy Delta S-2D of fatty-alcohol monolayers on water. Delta S-2D is found to be 4(kB)/mol lower than in the bulk. Because of the role of the conformational entropy, the melting transition is discontinuous for long chains, but tends to be...

  13. Drag Moderation by the Melting of an Ice Surface in Contact with Water

    KAUST Repository

    Vakarelski, Ivan Uriev

    2015-07-24

    We report measurements of the effects of a melting ice surface on the hydrodynamic drag of ice-shell-metal-core spheres free falling in water at a Reynolds of number Re∼2×104–3×105 and demonstrate that the melting surface induces the early onset of the drag crisis, thus reducing the hydrodynamic drag by more than 50%. Direct visualization of the flow pattern demonstrates the key role of surface melting. Our observations support the hypothesis that the drag reduction is due to the disturbance of the viscous boundary layer by the mass transfer from the melting ice surface.

  14. Drag Moderation by the Melting of an Ice Surface in Contact with Water

    KAUST Repository

    Vakarelski, Ivan Uriev; Chan, Derek Y.  C.; Thoroddsen, Sigurdur T

    2015-01-01

    We report measurements of the effects of a melting ice surface on the hydrodynamic drag of ice-shell-metal-core spheres free falling in water at a Reynolds of number Re∼2×104–3×105 and demonstrate that the melting surface induces the early onset of the drag crisis, thus reducing the hydrodynamic drag by more than 50%. Direct visualization of the flow pattern demonstrates the key role of surface melting. Our observations support the hypothesis that the drag reduction is due to the disturbance of the viscous boundary layer by the mass transfer from the melting ice surface.

  15. Study on mitigation of in-vessel release of fission products in severe accidents of PWR

    International Nuclear Information System (INIS)

    Huang, G.F.; Tong, L.L.; Li, J.X.; Cao, X.W.

    2010-01-01

    Research highlights: → In-vessel release of fission products in severe accidents for 600 MW PWR is analyzed. → Mitigation effect of primary feed-and-bleed on in-vessel release is investigated. → Mitigation effect of secondary feed-and-bleed on in-vessel release is studied. → Mitigation effect of ex-vessel cooling on in-vessel release is evaluated. - Abstract: During the severe accidents in a nuclear power plant, large amounts of fission products release with accident progression, including in-vessel and ex-vessel release. Mitigation of fission products release is demanded for alleviating radiological consequence in severe accidents. Mitigation countermeasures to in-vessel release are studied for Chinese 600 MW pressurized water reactor (PWR), including feed-and-bleed in primary circuit, feed-and-bleed in secondary circuit and ex-vessel cooling. SBO, LOFW, SBLOCA and LBLOCA are selected as typical severe accident sequences. Based on the evaluation of in-vessel release with different startup time of countermeasure, and the coupling relationship between thermohydraulics and in-vessel release of fission products, some results are achieved. Feed-and-bleed in primary circuit is an effective countermeasure to mitigate in-vessel release of fission products, and earlier startup time of countermeasure is more feasible. Feed-and-bleed in secondary circuit is also an effective countermeasure to mitigate in-vessel release for most severe accident sequences that can cease core melt progression, e.g. SBO, LOFW and SBLOCA. Ex-vessel cooling has no mitigation effect on in-vessel release owing to inevitable core melt and relocation.

  16. FARO tests corium-melt cooling in water pool: Roles of melt superheat and sintering in sediment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Gisuk [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States); Kaviany, Massoud [Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Moriyama, Kiyofumi [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Hwang, Byoungcheol; Lee, Mooneon; Kim, Eunho; Park, Jin Ho [Division of Advance Nuclear Engineering, POSTECH, Pohang, Gyeongbuk 790-784 (Korea, Republic of); Nasersharifi, Yahya [Department of Mechanical Engineering, Wichita State University, Wichita, KS 67260 (United States)

    2016-08-15

    Highlights: • The numerical approach for FARO experimental data is suggested. • The cooling mechanism of ex-vessel corium is suggested. • The predicted minimum pool depth for no cake formation is suggested. - Abstract: The FARO tests have aimed at understanding an important severe accident mitigation action in a light water reactor when the accident progresses from the reactor pressure vessel boundary. These tests have aimed to measure the coolability of a molten core material (corium) gravity dispersed as jet into a water pool, quantifying the loose particle diameter distribution and fraction converted to cake under range of initial melt superheat and pool temperature and depth. Under complete hydrodynamic breakup of corium and consequent sedimentation in the pool, the initially superheated corium can result in debris bed consisting of discrete solid particles (loose debris) and/or a solid cake at the bottom of the pool. The success of the debris bed coolability requires cooling of the cake, and this is controlled by the large internal resistance. We postulate that the corium cake forms when there is a remelting part in the sediment. We show that even though a solid shell forms around the melt particles transiting in the water pool due to film-boiling heat transfer, the superheated melt allows remelting of the large particles in the sediment (depending on the water temperature and the transit time) using the COOLAP (Coolability Analysis with Parametric fuel-cooant interaction models) code. With this remelting and its liquid-phase sintering of the non-remelted particles, we predict the fraction of the melt particles converting to a cake through liquid sintering. Our predictions are in good agreement with the existing results of the FARO experiments. We address only those experiments with pool depths sufficient/exceeding the length required for complete breakup of the molten jet. Our analysis of the fate of molten corium aimed at devising the effective

  17. Experimental investigation of ice and snow melting process on pavement utilizing geothermal tail water

    International Nuclear Information System (INIS)

    Wang Huajun; Zhao Jun; Chen Zhihao

    2008-01-01

    Road ice and snow melting based on low temperature geothermal tail water is of significance to realize energy cascading utilization. A small scale ice and snow melting system is built in this work. Experiments of dynamic melting processes of crushed ice, solid ice, artificial snow and natural snow are conducted on concrete pavement. The results show that the melting process of ice and snow includes three phases: a starting period, a linear period and an accelerated period. The critical value of the snow free area ratio between the linear period and the accelerated period is about 0.6. The physical properties of ice and snow, linked with ambient conditions, have an obvious effect on the melting process. The difference of melting velocity and melting time between ice and snow is compared. To reduce energy consumption, the formation of ice on roads should be avoided if possible. The idling process is an effective pathway to improve the performance of melting systems. It is feasible to utilize geothermal tail water of about 40 deg. C for melting ice and snow on winter roads, and it is unnecessary to keep too high fluid temperatures during the practical design and applications. Besides, with the exception of solid ice, the density and porosity of snow and ice tend to be decreasing and increasing, respectively, as the ambient temperature decreases

  18. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 4. Numerical analysis of 1/10 scaled water experiment with the AQUA code

    International Nuclear Information System (INIS)

    Muramatu, Toshiharu; Yamaguchi, Akira

    2004-01-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)

  19. 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

  20. On the influence of water subcooling and melt jet parameters on debris formation

    Energy Technology Data Exchange (ETDEWEB)

    Manickam, Louis, E-mail: louis@safety.sci.kth.se; Kudinov, Pavel; Ma, Weimin; Bechta, Sevostian; Grishchenko, Dmitry

    2016-12-01

    Highlights: • Melt and water configuration effects on debris formation is studied experimentally. • Melt superheat and water subcooling are most influential compared to jet size. • Melt-water configuration and material properties influence particle fracture rate. • Results are compared with large scale experiments to study effect of spatial scales. - Abstract: Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-Jet facility by discharging binary-oxide mixtures of WO{sub 3}–Bi{sub 2}O{sub 3} and WO{sub 3}–ZrO{sub 2} into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased subcooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

  1. Melt water interaction tests. PREMIX tests PM10 and PM11

    Energy Technology Data Exchange (ETDEWEB)

    Kaiser, A.; Schuetz, W.; Will, H. [Forschungszentrum Karlsruhe Inst. fuer Reaktorsicherheit, Karlsruhe (Germany)

    1998-01-01

    A series of experiments is being performed in the PREMIX test facility in which the mixing behaviour is investigated of a hot alumina melt discharged into water. The major parameters have been: the melt mass, the number of nozzles, the distance between the nozzle and the water, and the depth of the water. The paper describes the last two tests in which 20 kg of melt were released through one and three nozzles, respectively, directly into the water whose depth was 500 mm. The melt penetration and the associated phenomena of mixing are described by means of high-speed films and various measurements. The steam production and, subsequently, the pressure increased markedly only after the melt had reached the bottom of the pool. Spreading of the melt across the bottom caused violent boiling in both tests. Whereas the boiling lasted for minutes in the single-jet test, a steam explosion occurred in the triple-jet test about one second after the start of melt penetration. (author)

  2. Investigation on Melt-Structure-Water Interactions (MSWI) during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Yang, Z.L.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Haraldsson, H.O.; Li, H.X.; Konovakhin, M.; Paladino, D.; Leung, W.H [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    1999-08-01

    This report is the final report for the work performed in 1998 in the research project Melt Structure Water Interactions (MSWI), under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The present report describes results of advanced analytical and experimental studies concerning melt-water-structure interactions during the course of a hypothetical severe core meltdown accident in a light water reactor (LWR). Emphasis has been placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Many of the investigations performed in support of this project have produced papers which have been published in the proceedings of technical meetings. A short summary of the results achieved in these papers is provided in this overview. Both experimental and analytical studies were performed to improve knowledge about phenomena of melt-structure-water interactions. We believe that significant technical advances have been achieved during the course of these studies. It was found that: the solidification has a strong effect on the drop deformation and breakup. Initially appearing at the drop surface and, later, thickening inwards, the solid crust layer dampens the instability waves on the drop surface and, therefore, hinders drop deformation and breakup. The drop thermal properties also affect the thermal behavior of the drop and, therefore, have impact on its deformation behavior. The jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters, e.g. the Weber number, but also on the melt physical properties, which change as the melt cools down from the liquidus to the solidus temperature. Additionally, the crust formed on the surface of the melt jet will also reduce the propensity

  3. Investigation on Melt-Structure-Water Interactions (MSWI) during severe accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Yang, Z.L.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Haraldsson, H.O.; Li, H.X.; Konovakhin, M.; Paladino, D.; Leung, W.H

    1999-08-01

    This report is the final report for the work performed in 1998 in the research project Melt Structure Water Interactions (MSWI), under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The present report describes results of advanced analytical and experimental studies concerning melt-water-structure interactions during the course of a hypothetical severe core meltdown accident in a light water reactor (LWR). Emphasis has been placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Many of the investigations performed in support of this project have produced papers which have been published in the proceedings of technical meetings. A short summary of the results achieved in these papers is provided in this overview. Both experimental and analytical studies were performed to improve knowledge about phenomena of melt-structure-water interactions. We believe that significant technical advances have been achieved during the course of these studies. It was found that: the solidification has a strong effect on the drop deformation and breakup. Initially appearing at the drop surface and, later, thickening inwards, the solid crust layer dampens the instability waves on the drop surface and, therefore, hinders drop deformation and breakup. The drop thermal properties also affect the thermal behavior of the drop and, therefore, have impact on its deformation behavior. The jet fragmentation process is a function of many related phenomena. The fragmentation rate depends not only on the traditional parameters, e.g. the Weber number, but also on the melt physical properties, which change as the melt cools down from the liquidus to the solidus temperature. Additionally, the crust formed on the surface of the melt jet will also reduce the propensity

  4. Experimental study of simulant melt stream-water thermal interaction in pool and narrow geometries

    International Nuclear Information System (INIS)

    Narayanan, K.S.; Jasmin Sudha, A.; Murthy, S.S.; Rao, E.H.V.M.; Lydia, G.; Das, S.K.; Harvey, J.; Kannan, S.E.

    2005-01-01

    Full text of publication follows: Small scale experiments were carried out to investigate the thermal interaction characteristics of a few kilograms of Sn Pb, Bi and Zn as hot melt, in the film boiling region of water in an attempt to simulate a coherent fuel coolant interaction during a postulated severe accident in a nuclear reactor. Melt stream solidification and detached debris generation were studied with different melt superheat up to 200 deg. C, at different coolant temperatures of 30 deg. C, 50 deg. C, 70 deg. C, 90 deg. C, in pool geometry and in long narrow coolant column. The material was heated in an Alumina crucible and poured through a hot stainless steel funnel with a nozzle diameter of 7.7 mm, into the coolant. A stainless steel plate was used to collect the solidified mass after the interaction. Temperature monitoring was done in the coolant column close to the melt stream. The melt stream movement inside the coolant was imaged using a video camera at 25 fps. Measured melt stream entry velocity was around 1.5 m/sec. For low melt superheat and low coolant temperature, solidified porous tree like structure extended from the collector plate up to the melt release point. For water temperature of 70 deg. C, the solidified bed height at the center was found to decrease with increase in the melt superheat up to 150 deg. C. Fragmentation was found to occur when the melt superheat exceeded 200 deg. C. Particle size distribution was obtained for the fragmented debris. In 1D geometry, with 50 deg. C superheat, columnar solidification was observed with no fine debris. The paper gives the details of the results obtained in the experiments and highlights the role of Rayleigh-Taylor, Kelvin-Helmholtz instabilities and the melt physical properties on the fragmentation kinetics. (authors)

  5. An effective convectivity model for simulation of in-vessel core melt progression in a boiling water reactor

    International Nuclear Information System (INIS)

    Tran, C.T.; Dinh, T.N.

    2007-01-01

    The present paper is concerned with development and application of a so-called Effective Convection Model (ECM), which aims to provide a detailed, mechanistic description of heat transfer processes in a BWR lower plenum. The ECM is a Computational Fluid Dynamics (CFD)-like tool which employs a simpler and more effective approach to compute heat transfer by solving only energy conservation equation instead of solving the full set of Navier-Stokes and energy equations by a CFD code. We implement the ECM in a CFD code (Fluent), with detailed description of the ECM development, implementation and validation. A dual approach is used to validate the ECM, namely validation against experimental data and against heat transfer results obtained by CFD predictions in the same geometries and conditions. Insights gained from CFD simulations are also used to improve ECM. The ECM capability as an effective tool to simulate heat transfer of an internally heated volume in 3-dimensional complex geometry is demonstrated through examples of heat transfer analysis in a BWR lower plenum being cooled by coolant flow in Control Rod Guide Tubes. Simulation results and key findings of this case are reported and discussed. (authors)

  6. Three-Phase Melting Curves in the Binary System of Carbon Dioxide and Water

    Science.gov (United States)

    Abramson, E. H.

    2017-10-01

    Invariant, three-phase melting curves, of ice VI in equilibrium with solid CO2, of ice VII in equilibrium with solid CO2, and of solid CO2 in simultaneous equilibrium with a majority aqueous and a majority CO2 fluid, were explored in the binary system of carbon dioxide and water. Diamond-anvil cells were used to develop pressures of 5 GPa. Water exhibits a large melting temperature depression (73°C less than its pure melting temperature of 253°C at 5 GPa) indicative of large concentrations of CO2 in the aqueous solution. The melting point of water-saturated CO2 does not show a measureable departure from that of the pure system at temperatures lower than ∼200°C and only 10°C at 5 GPa (from 327°C).

  7. Experimental investigation of 150-KG-scale corium melt jet quenching in water

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Hohmann, H.

    1995-09-01

    This paper compares and discusses the results of two large scale FARO quenching tests known as L-11 and L-14, which involved, respectively, 151 kg of W% 76.7 UO{sub 2} + 19.2 ZrO{sub 2} + 4.1 Zr and 125 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melts poured into 600-kg, 2-m-depth water at saturation at 5.0 MPa. The results are further compared with those of two previous tests performed using a pure oxidic melt, respectively 18 and 44 kg of W% 80 UO{sub 2} + 20 ZrO{sub 2} melt quenched in 1-m-depth water at saturation at 5.0 MPa. In all the tests, significant breakup and quenching took place during the melt fall through the water. No steam explosion occurred. In the tests performed with a pure oxide UO{sub 2}-ZrO{sub 2} melt, part of the corium (from 1/6 to 1/3) did not breakup and reached the bottom plate still molten whatever the water depth was. Test L-11 data suggest that full oxidation and complete breakup of the melt occurred during the melt fall through the water. A proportion of 64% of the total energy content of the melt was released to the water during this phase ({approximately}1.5 s), against 44% for L-14. The maximum temperature increase of the bottom plate was 330 K (L-14). The mean particle size of the debris ranged between 2.5 and 4.8mm.

  8. Investigation of the Potential for In-Vessel Melt Retention in the Lower Head of a BWR by Cooling through the Control Rod Guide Tubes. APRl 4, Stage 2 Report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Jasiulevicius, A.; Konovalikhin, M.

    2004-01-01

    This report describes the experiments performed at the Division, investigating the coolability potential offered by the Control Rod Guide Tubes (CRGTs), which are present in large numbers in the lower head of a BWR and there is a water flow circuit in each one of them. This investigation is related to the overall goal of retaining the core melt in the lower head of a BWR during a postulated severe accident, through accident management procedures, or strategy. The experiments were performed in two facilities, i.e. POMECO (Porous MEdium COolability) and COMECO (Core MElt COolability), respectively, for investigating the coolability when the core material is in the form of a particulate debris bed and when it is in the form of a melt. The POMECO facility employed a sand bed heated electrically to heating levels of up to 1 MW/m 3 and experiments performed in that facility obtained the enhancement in the dryout heat flux and in the quench velocity due to presence of a CRGT, with, and without, water flow in it. The COMECO facility employed a simulant material melt pool heated electrically to power levels of = 1.3 MW/m 3 and the experiments in it also determined the enhancement in the heat removal from the melt pool that could be obtained by the presence of a CRGT, with, or without water flow in it. In each of the experiments in these facilities, the scaling employed was of a unit cell of core material around a prototypic geometry CRGT with the prototypic decay heat input. The experimental results showed that a CRGT is able to offer a substantial additional potential for coolability of particulate and melt material in the lower head of a BWR. Analysis of the data obtained in the set of experiments performed lead to the following results for the heat flux through the CRGT: - for a water filled particulate debris bed: - 40 kW/m 2 ; - for a day hot particulate debris bed: - 150 kW/m 2 ; - for a melt pool with a crust formed on the CRGT surface: -350 kW/m 2 . It is

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

    International Nuclear Information System (INIS)

    Ghyym, Seong H.

    1998-01-01

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

  10. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    Energy Technology Data Exchange (ETDEWEB)

    M. Anderson; M. Corradini; K.Y. Bank; R. Bonazza; D. Cho

    2005-04-26

    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 this 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 in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.

  11. Description of the heating and expansion process of a water drop enclosed in a hot melt

    International Nuclear Information System (INIS)

    Froehlich, G.; Berg, E. von.

    1985-11-01

    In the present study a simple model for the description of the heating- and expansion-process of a water drop enclosed in hot melt is developed. The model is valid between the first contact of melt and water up to the beginning of evaporation. A possible superheating by retardation of ebullition is disregarded. The balance equations for energy, mass and momentum as well as the equation of state are integrated over the radial space coordinate in both media using appropriate profiles of temperature, pressure and velocity. Thereby a system of coupled ordinary differential equations is formed for the variables of the model which are now time dependent only. The equations are solved numerically by means of a FORTRAN-program. The influence of parameters (melt-temperature, heat-transfer-coefficient between melt and water as well as drop radius) are studied. It is shown that always very rapidly a vapor-layer forms around the water drop, while the inner part of the drop did not yet 'notice' anything of the heating process. An approximation formula for the time-transfer-coefficients between melt and water. Due to this approximation, the time up to incipience of evaporation grows proportional to the drop radius, which means that in the frame of the present model even small droplets won't evaporate as a whole instantaneously. (orig.) [de

  12. Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

    International Nuclear Information System (INIS)

    Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R.; Cho, D.

    2005-01-01

    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 this 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 in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications

  13. Phenomenological studies on melt-structure-water interactions (MSWI) during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Yang, Z.L.; Haraldsson, H.O.; Nourgaliev, R.R.; Konovalikhin, M.; Paladino, D.; Gubaidullin, A.A.; Kolb, G.; Theerthan, A. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety

    2000-05-01

    This is the annual report for the work performed in 1999 in the research project Melt-Structure-Water Interactions During Severe Accidents in LWRs, under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The emphasis of the work is placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. We believe that significant technical advances have been achieved during the course of these studies. It was found that: The coolant temperature has significant influence on the characteristics of debris fragments produced from the breakup of an oxidic melt jet. At low subcooling the fragments are relatively large and irregular compared to the smaller particles produced at high subcooling. The melt jet density has considerable effect on the fragment size produced. As the melt density increases the fragment size becomes smaller. The mass mean size of the debris changes proportionally to the square root of the coolant to melt density ratio. The melt superheat has little effect on the debris particle size distribution produced during the melt jet fragmentation. The impingement velocity of the jet has significant impact on the fragmentation process. At lower jet velocity the melt fragments agglomerate and form a cake of large size debris. When the jet velocity is increased more complete fragmentation is obtained. The scaling methodology for melt spreading, developed during 1998, has been further validated against almost all of the spreading experimental data available so far. Experimental results for the dryout heat flux of homogeneous particulate debris beds with top flooding compare well with the Lipinski correlation. For the stratified particle beds, the fine particle layer resting on the top of another particle layer dominates the dryout processes

  14. Phenomenological studies on melt-structure-water interactions (MSWI) during severe accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Yang, Z.L.; Haraldsson, H.O.; Nourgaliev, R.R.; Konovalikhin, M.; Paladino, D.; Gubaidullin, A.A.; Kolb, G.; Theerthan, A.

    2000-05-01

    This is the annual report for the work performed in 1999 in the research project Melt-Structure-Water Interactions During Severe Accidents in LWRs, under the auspices of the APRI Project, jointly funded by SKI, HSK, USNRC and the Swedish and Finnish power companies. The emphasis of the work is placed on phenomena and properties which govern the fragmentation and breakup of melt jets and droplets, melt spreading and coolability, and thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. We believe that significant technical advances have been achieved during the course of these studies. It was found that: The coolant temperature has significant influence on the characteristics of debris fragments produced from the breakup of an oxidic melt jet. At low subcooling the fragments are relatively large and irregular compared to the smaller particles produced at high subcooling. The melt jet density has considerable effect on the fragment size produced. As the melt density increases the fragment size becomes smaller. The mass mean size of the debris changes proportionally to the square root of the coolant to melt density ratio. The melt superheat has little effect on the debris particle size distribution produced during the melt jet fragmentation. The impingement velocity of the jet has significant impact on the fragmentation process. At lower jet velocity the melt fragments agglomerate and form a cake of large size debris. When the jet velocity is increased more complete fragmentation is obtained. The scaling methodology for melt spreading, developed during 1998, has been further validated against almost all of the spreading experimental data available so far. Experimental results for the dryout heat flux of homogeneous particulate debris beds with top flooding compare well with the Lipinski correlation. For the stratified particle beds, the fine particle layer resting on the top of another particle layer dominates the dryout processes

  15. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

    2016-10-15

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

  16. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

    2016-01-01

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

  17. An Overview Of The ITER In-Vessel Coil Systems

    International Nuclear Information System (INIS)

    Heitzenroeder, P.J.; Brooks, A.W.; Chrzanowski, J.H.; Dahlgren, F.; Hawryluk, R.J.; Loesser, G.D.; Neumeyer, C.; Mansfield, C.; Smith, J.P.; Schaffer, M.; Humphreys, D.; Cordier, J.J.; Campbell, D.; Johnson, G.A.; Martin, A.; Rebut, P.H.; Tao, J.O.; Fogarty, P.J.; Nelson, B.E.; Reed, R.P.

    2009-01-01

    ELM mitigation is of particular importance in ITER in order to prevent rapid erosion or melting of the divertor surface, with the consequent risk of water leaks, increased plasma impurity content and disruptivity. Exploitable 'natural' small or no ELM regimes might yet be found which extrapolate to ITER but this cannot be depended upon. Resonant Magnetic Perturbation has been added to pellet pacing as a tool for ITER to mitigate ELMs. Both are required, since neither method is fully developed and much work remains to be done. In addition, in-vessel coils enable vertical stabilization and RWM control. For these reasons, in-vessel coils (IVCs) are being designed for ITER to provide control of Edge Localized Modes (ELMs) in addition to providing control of moderately unstable resistive wall modes (RWMs) and the vertical stability (VS) of the plasma.

  18. Human activities and its Responses to Glacier Melt Water Over Tarim River Basin

    Science.gov (United States)

    He, Hai; Zhou, Shenbei; Bai, Minghao

    2017-04-01

    Tarim River Basin lies in the south area of Xinjiang Uygur Autonomous Region, the north-west area of China. It is the longest inland river of China. Being far away from ocean and embraced by high mountains, Tarim River Basin is the typical arid region in the world. The intensity of human activities increased rapidly in Tarim River Basin since 1980's and water resources lacking is the major issue restricting the development of social economy. The glacier melt water plays an important role for the regional social and economic development, and it accounts for 40% of mountain-pass runoff. It is a fragile mutual-dependent relationship between local sustainable development and runoff. Under the background of global change glacier melt water process has also changed especially in the arid and semi-arid region. Due to climate change, glacier in Tarim River Basin has melted in an observed way since 1980s, together with increasing trend of annual rainfall and virgin flow in mountain basins. Correspondingly, human activity gets more frequent since 1970s, resulting into the obvious fragile mutual-dependent relationship between basin runoff and water use amount. Through an analysis of meteorological, hydrological and geographical observation data from 1985 to 2015, this thesis make a multi-factor variance analysis of population, cultivation area, industrial development and runoff in upstream and mid-stream of Tarim River under changing conditions. Furthermore, the regulation function of natural factors and water demand management factors on relationship between runoff and water using amount are discussed, including temperature, rainfall, and evaporation, water conservation technology and soil-water exploitation administrative institutions. It concludes that: first, increase in glacier runoff, rainfall amount, and virgin flow haven't notably relieved ecological issue in Tarim River Basin, and even has promoted water use behaviour in different flowing areas and noticeably reduced

  19. Water- and Boron-Rich Melt Inclusions in Quartz from the Malkhan Pegmatite, Transbaikalia, Russia

    Directory of Open Access Journals (Sweden)

    Elena Badanina

    2012-11-01

    Full Text Available In this paper we show that the pegmatite-forming processes responsible for the formation of the Malkhan pegmatites started at magmatic temperatures around 720 °C. The primary melts or supercritical fluids were very water- and boron-rich (maximum values of about 10% (g/g B2O3 and over the temperature interval from 720 to 600 °C formed a pseudobinary solvus, indicated by the coexistence of two types of primary melt inclusions (type-A and type-B representing a pair of conjugate melts. Due to the high water and boron concentration the pegmatite-forming melts are metastable and can be characterized either as genuine melts or silicate-rich fluids. This statement is underscored by Raman spectroscopic studies. This study suggested that the gel state proposed by some authors cannot represent the main stage of the pegmatite-forming processes in the Malkhan pegmatites, and probably in all others. However there are points in the evolution of the pegmatites where the gel- or gel-like state has left traces in form of real gel inclusions in some mineral in the Malkhan pegmatite, however only in a late, fluid dominated stage.

  20. Modeling radar backscattering from melting snowflakes using spheroids with nonuniform distribution of water

    International Nuclear Information System (INIS)

    Tyynelä, Jani; Leinonen, Jussi; Moisseev, Dmitri; Nousiainen, Timo; Lerber, Annakaisa von

    2014-01-01

    In a number of studies it is reported that at the early stages, melting of aggregate snowflakes is enhanced at lower parts. In this paper, the manifestation of the resulting nonuniform distribution of water is studied for radar backscattering cross sections at C, Ku, Ka and W bands. The melting particles are described as spheroids with a mixture of water and air at the bottom part of the particle and a mixture of ice and air at the upper part. The radar backscattering is modeled using the discrete-dipole approximation in a horizontally pointing geometry. The results are compared to the T-matrix method, Mie theory, and the Rayleigh approximation using the Maxwell Garnett mixing formula. We find that the differential reflectivity and the linear depolarization ratio show systematic differences between the discrete-dipole approximation and the T-matrix method, but that the differences are relatively small. The horizontal cross sections show only small differences between the methods with the aspect ratio and the presence of resonance peaks having a larger effect on it than the nonuniform distribution of water. Overall, the effect of anisotropic distribution of water, reported for early stages of melting, is not significant for radar observations at the studied frequencies. -- Highlights: • We model backscattering from spheroidal melting snowflakes at C, Ku, Ka, and W bands. • We study the effect of anisotropic distribution of meltwater in the snow particles. • We find systematic, but relatively small differences for the backscattering properties. • We find that the aspect ratio and resonance peaks have a bigger effect than anisotropic distribution of water. • Anisotropic distribution of water is not significant for radar observations at early stages of melting

  1. Experiments on interactions between zirconium-containing melt and water (ZREX). Hydrogen generation and chemical augmentation of energetics

    Energy Technology Data Exchange (ETDEWEB)

    Cho, D.H.; Armstrong, D.R.; Gunther, W.H. [Argonne National Lab., IL (United States); Basu, S.

    1998-01-01

    The results of the first data series of experiments on interactions between zirconium-containing melt and water are described. These experiments involved dropping 1-kg batches of pure zirconium or zirconium-zirconium dioxide mixture melt into a column of water. A total of nine tests were conducted, including four with pure zirconium melt and five with Zr-ZrO{sub 2} mixture melt. Explosions took place only in those tests which were externally triggered. While the extent of zirconium oxidation in the triggered experiments was quite extensive, the estimated explosion energetics were found to be very small compared to the combined thermal and chemical energy available. (author)

  2. Air-sea flux of CO2 in arctic coastal waters influenced by glacial melt water and sea ice

    DEFF Research Database (Denmark)

    Sejr, Mikael Kristian; Krause-Jensen, Dorte; Rysgaard, Søren

    2011-01-01

    Annual air–sea exchange ofCO2 inYoung Sound,NEGreenlandwas estimated using pCO2 surface-water measurements during summer (2006–2009) and during an ice-covered winter 2008. All surface pCO2 values were below atmospheric levels indicating an uptake of atmospheric CO2. During sea ice formation...... and thereby efficiently blocked air–sea CO2 exchange. During sea ice melt, dissolution of CaCO3 combined with primary production and strong stratification of the water column acted to lower surface-water pCO2 levels in the fjord. Also, a large input of glacial melt water containing geochemically reactive...... year-to-year variation in annual gas exchange....

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-06-01

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

  4. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.; Kenton, M.A.

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations

  5. Water speciation in sodium silicate glasses (quenched melts): A comprehensive NMR study

    Science.gov (United States)

    Xue, X.; Kanzaki, M.; Eguchi, J.

    2012-12-01

    Dissolution mechanism of water is an important factor governing how the dissolved water affects the physical and thermodynamic properties of silicate melts and glasses. Our previous studies have demonstrated that 1H MAS NMR in combination with 29Si-1H and 27Al-1H double-resonance NMR experiments is an effective approach for unambiguously differentiating and quantifying different water species in quenched silicate melts (glasses). Several contrasting dissolution mechanisms have been revealed depending on the melt composition: for relatively polymerized melts, the formation of SiOH/AlOH species (plus molecular H2O) and depolymerization of the network structure dominate; whereas for depolymerized Ca-Mg silicate melts, free OH (e.g. MgOH) become increasingly important (cf. [1]). The proportion of free OH species has been shown to decrease with both increasing melt polymerization (silica content) and decreasing field strength of the network modifying cations (from Mg to Ca). Our previous 1H and 29Si MAS NMR results for hydrous Na silicate glasses of limited compositions (Na2Si4O9 and Na2Si2O5) were consistent with negligible free OH (NaOH) species and depolymerizing effect of water dissolution [2]. On the other hand, there were also other studies that proposed the presence of significant NaOH species in hydrous glasses near the Na2Si2O5 composition. The purpose of this study is apply the approach of combined 1H MAS NMR and double-resonance (29Si-1H and 23Na-1H) NMR to gain unambiguous evidence for the OH speciation in Na silicate glasses (melts) as a function of composition. Hydrous Na silicate glasses containing mostly ≤ 1 wt% H2O for a range of Na/Si ratios from 0.33 to 1.33 have been synthesized by rapidly quenching melts either at 0.2 GPa using an internally heated gas pressure vessel or at 1 GPa using a piston cylinder high-pressure apparatus. NMR spectra have been acquired using a 9.4 T Varian Unity-Inova spectrometer. The 29Si and 1H chemical shifts are

  6. Diffusion of Water through Olivine and Clinopyroxene: Implications for Melt Inclusion Fidelity

    Science.gov (United States)

    Plank, T. A.; Lloyd, A. S.; Ferriss, E.

    2016-12-01

    The maximum H2O concentrations measured in olivine-hosted melt inclusions (MIs) from arc tephra fall within a narrow range of 3-5 wt%. A major question is whether this reflects parental water concentrations or diffusive exchange through the host crystal during storage and ascent. Laboratory experiments have shown that water can diffuse through 500 micron olivine in minutes to days at 1100°C. We have tested these predictions with a natural experiment using volatile (H2O, CO2, S) diffusion along melt embayments to constrain ascent rates during the 1974 eruption of Volcan Fuego to 5-8 minutes from 7 km depth [1]. Thus, olivine-hosted MIs may move from their storage region to the surface during some eruptions rapidly enough to retain almost all of their original water. Only the smallest MIs (500 microns) and large melt inclusions (>50 microns), and 4) rapid post-eruptive cooling (< 1min, clast sizes < 1 cm). The rapid diffusion of H through olivine and cpx presents a challenge to MI fidelity, but not necessarily if the above conditions are met. [1] Lloyd et al., 2014, JVGR. [2] Ferriss et al., 2016, AmMin.

  7. Measurement of the Spectral Absorption of Liquid Water in Melting Snow With an Imaging Spectrometer

    Science.gov (United States)

    Green, Robert O.; Dozier, Jeff

    1995-01-01

    Melting of the snowpack is a critical parameter that drives aspects of the hydrology in regions of the Earth where snow accumulates seasonally. New techniques for measurement of snow melt over regional scales offer the potential to improve monitoring and modeling of snow-driven hydrological processes. In this paper we present the results of measuring the spectral absorption of liquid water in a melting snowpack with the Airborne Visible/Infrared Imaging Spectrometer (AVIRIS). AVIRIS data were acquired over Mammoth Mountain, in east central California on 21 May 1994 at 18:35 UTC. The air temperature at 2926 m on Mammoth Mountain at site A was measured at 15-minute intervals during the day preceding the AVIRIS data acquisition. At this elevation. the air temperature did not drop below freezing the night of the May 20 and had risen to 6 degrees Celsius by the time of the overflight on May 21. These temperature conditions support the presence of melting snow at the surface as the AVIRIS data were acquired.

  8. Rapid changes in surface water carbonate chemistry during Antarctic sea ice melt

    Science.gov (United States)

    Jones, Elizabeth M.; Bakker, Dorothee C. E.; Venables, Hugh J.; Whitehouse, Michael J.; Korb, Rebecca E.; Watson, Andrew J.

    2010-11-01

    ABSTRACT The effect of sea ice melt on the carbonate chemistry of surface waters in the Weddell-Scotia Confluence, Southern Ocean, was investigated during January 2008. Contrasting concentrations of dissolved inorganic carbon (DIC), total alkalinity (TA) and the fugacity of carbon dioxide (fCO2) were observed in and around the receding sea ice edge. The precipitation of carbonate minerals such as ikaite (CaCO3.6H2O) in sea ice brine has the net effect of decreasing DIC and TA and increasing the fCO2 in the brine. Deficits in DIC up to 12 +/- 3 μmol kg-1 in the marginal ice zone (MIZ) were consistent with the release of DIC-poor brines to surface waters during sea ice melt. Biological utilization of carbon was the dominant processes and accounted for 41 +/- 1 μmol kg-1 of the summer DIC deficit. The data suggest that the combined effects of biological carbon uptake and the precipitation of carbonates created substantial undersaturation in fCO2 of 95 μatm in the MIZ during summer sea ice melt. Further work is required to improve the understanding of ikaite chemistry in Antarctic sea ice and its importance for the sea ice carbon pump.

  9. Decompression-induced melting of ice IV and the liquid-liquid transition in water

    Science.gov (United States)

    Mishima, Osamu; Stanley, H. Eugene

    1998-03-01

    Although liquid water has been the focus of intensive research for over 100 years, a coherent physical picture that unifies all of the known anomalies of this liquid, is still lacking. Some of these anomalies occur in the supercooled region, and have been rationalized on the grounds of a possible retracing of the liquid-gas spinodal (metastability limit) line into the supercooled liquid region, or alternatively the presence of a line of first-order liquid-liquid phase transitions in this region which ends in a critical point,. But these ideas remain untested experimentally, in part because supercooled water can be probed only above the homogeneous nucleation temperature TH at which water spontaneously crystallizes. Here we report an experimental approach that is not restricted by the barrier imposed by TH, involving measurement of the decompression-induced melting curves of several high-pressure phases of ice in small emulsified droplets. We find that the melting curve for ice IV seems to undergo a discontinuity at precisely the location proposed for the line of liquid-liquid phase transitions. This is consistent with, but does not prove, the coexistence of two different phases of (supercooled) liquid water. From the experimental data we calculate a possible Gibbs potential surface and a corresponding equation of state for water, from the forms of which we estimate the coordinates of the liquid-liquid critical point to be at pressure Pc ~ 0.1GPa and temperature Tc ~ 220K.

  10. Densification behavior of gas and water atomized 316L stainless steel powder during selective laser melting

    Science.gov (United States)

    Li, Ruidi; Shi, Yusheng; Wang, Zhigang; Wang, Li; Liu, Jinhui; Jiang, Wei

    2010-04-01

    The densification during selective laser melting (SLM) process is an important factor determining the final application of SLM-part. In the present work, the densifications under different processing conditions were investigated and the densification mechanisms were elucidated. It was found that the higher laser power, lower scan speed, narrower hatch spacing and thinner layer thickness could enable a much smoother melting surface and consequently a higher densification. The gas atomized powder possessed better densification than water atomized powder, due to the lower oxygen content and higher packing density of gas atomized powder. A large number of regular-shaped pores can be generated at a wider hatch spacing, even if the scanning track is continuous and wetted very well. The densification mechanisms were addressed and the methods for building dense metal parts were also proposed as follows: inhibiting the balling phenomenon, increasing the overlap ratio of scanning tracks and reducing the micro-cracks.

  11. Oxygen exchange and ice melt measured at the ice-water interface by eddy correlation

    DEFF Research Database (Denmark)

    Long, M.H.; Koopmans, D.; Berg, P.

    2012-01-01

    heterotrophic with a daily gross primary production of 0.69 mmol O2 mĝ̂'2 dĝ̂'1 and a respiration rate of ĝ̂'2.13 mmol O2 mĝ̂'2 dĝ̂'1 leading to a net ecosystem metabolism of ĝ̂'1.45 mmol O2 mĝ̂'2 dĝ̂'1. This application of the eddy correlation technique produced high temporal resolution O2 fluxes and ice melt......This study examined fluxes across the ice-water interface utilizing the eddy correlation technique. Temperature eddy correlation systems were used to determine rates of ice melting and freezing, and O2 eddy correlation systems were used to examine O2 exchange rates driven by biological and physical...

  12. In-vessel tritium

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  13. Verification of IVA5 computer code for melt-water interaction analysis. Pt. 2. Three-phase flows with melt fragmentation

    International Nuclear Information System (INIS)

    Kolev, N.I.

    1999-01-01

    In order to qualify IVA5 for applications in the field of the melt-water interactions in nuclear reactor safety, we analyzed the achievable accuracy by predicting phenomena that are within this class. Comparison with FARO and PREMIX experiments characterized with dynamic fragmentation of the participating materials together With the comparison with the variety of experiments documented in part 1 of this work qualified IVA5 as a code representing the state-of-the-art in the field of the multiphase flows. The code is capable of predicting multi-phase flow behavior in complicated 3D geometries and industrial networks. The code is able to predict melt-water interaction in well quantified uncertainty region. Reducing the uncertainty band needs future sophistication in the directions specified in this work. (author)

  14. Entrapment investigations of water-droplet behavior in a hot tin melt with varying discharge velocities and orifices

    International Nuclear Information System (INIS)

    Froehlich, G.; Mueller, K.

    1983-10-01

    Experiments were performed in which water was pressed through a thermally isolated tube into a clyindrical crucible (diameter 5 cm, height 7,5 cm both measured inside) filled with molten tin (600 K). The diameter of the circular water outlet was varied from 0.5 up to 10 mm and the discharge velocity of the water was in the range of 0.05 up to 20 m/s. In the tin melt the water divides into single drops, which emerged on the melt surface, if an interaction between water and tin melt did not occur. The probability for an interaction increased in experiments with higher discharge velocities of the water and smaller diameters of the water outlet. In experiments with discharge velocities ≥ 5 m/s and outlet diameters ≤ 2 mm one or more interactions occured in each case. At these interactions of water drops entrapped in the tin melt (called entrapment interactions) a portion of the melt was ejected from the crucible. The moment of the interaction and the pulse of the force toward the crucible bottom were recorded. (orig.) [de

  15. Experimental Measurement of Frozen and Partially Melted Water Droplet Impact Dynamics

    Science.gov (United States)

    Palacios, Jose; Yan, Sihong; Tan, Jason; Kreeger, Richard E.

    2014-01-01

    High-speed video of single frozen water droplets impacting a surface was acquired. The droplets diameter ranged from 0.4 mm to 0.9 mm and impacted at velocities ranging from 140 m/sec to 309 m/sec. The techniques used to freeze the droplets and launch the particles against the surfaces is described in this paper. High-speed video was used to quantify the ice accretion area to the surface for varying impact angles (30 deg, 45 deg, 60 deg), impacting velocities, and break-up angles. An oxygen /acetylene cross-flow flame used to ensure partial melting of the traveling frozen droplets is also discussed. A linear relationship between impact angle and ice accretion is identified for fully frozen particles. The slope of the relationship is affected by impact speed. Perpendicular impacts, i.e. 30 deg, exhibited small differences in ice accretion for varying velocities, while an increase of 60% in velocity from 161 m/sec to 259 m/sec, provided an increase on ice accretion area of 96% at an impact angle of 60 deg. The increase accretion area highlights the importance of impact angle and velocity on the ice accretion process of ice crystals. It was experimentally observed that partial melting was not required for ice accretion at the tested velocities when high impact angles were used (45 and 60 deg). Partially melted droplets doubled the ice accretion areas on the impacting surface when 0.0023 Joules were applied to the particle. The partially melted state of the droplets and a method to quantify the percentage increase in ice accretion area is also described in the paper.

  16. Attenuation and Velocity Structure in Spain and Morocco: Distinguishing Between Water, Temperature, and Partial Melt

    Science.gov (United States)

    Bezada, M. J.; Humphreys, E.

    2014-12-01

    Temperature, melt fraction, and water content affect seismic velocity and attenuation differently. Both are sensitive to temperature, but velocity is more sensitive to melt fraction and attenuation is thought to be more sensitive to water content. For these reasons, combining attenuation measurements with tomographic imaging of velocity structure can help untangle these fields and better resolve lithospheric structure and physical state. We map variations in attenuation beneath Spain and northern Morocco using teleseismic data generated by more than a dozen teleseismic deep-focus earthquakes recorded on a dense array of stations. For each event, we first estimate the source from the best quality recordings. We then apply an attenuation operator to the source estimate, using a range of t* values, to match the record at each station. We invert for a smooth map of t* from the ensemble of measurements. The spatial patterns in t* correlate very well with the tectonic domains in Spain and Morocco. In particular, areas in Spain that resisted deformation during the Variscan and Alpine orogenies produce very little attenuation. Comparing the attenuation map with seismic velocity structure we find that, in Morocco, some areas with strong low-velocity anomalies and recent volcanism do not cause high attenuation. These observations suggest that water content is a more likely cause for seismic attenuation in the study area than temperature, and that the non-attenuative low-velocity anomalies in Morocco are produced by partial mel.

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

    International Nuclear Information System (INIS)

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

    2001-03-01

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

  18. Water Content of Earth's Continental Mantle Is Controlled by the Circulation of Fluids or Melts

    Science.gov (United States)

    Peslier, Anne; Woodland, Alan B.; Bell, David R.; Lazarov, Marina; Lapen, Thomas J.

    2014-01-01

    A key mission of the ARES Directorate at JSC is to constrain models of the formation and geological history of terrestrial planets. Water is a crucial parameter to be measured with the aim to determine its amount and distribution in the interior of Earth, Mars, and the Moon. Most of that "water" is not liquid water per se, but rather hydrogen dissolved as a trace element in the minerals of the rocks at depth. Even so, the middle layer of differentiated planets, the mantle, occupies such a large volume and mass of each planet that when it is added at the planetary scale, oceans worth of water could be stored in its interior. The mantle is where magmas originate. Moreover, on Earth, the mantle is where the boundary between tectonic plates and the underlying asthenosphere is located. Even if mantle rocks in Earth typically contain less than 200 ppm H2O, such small quantities have tremendous influence on how easily they melt (i.e., the more water there is, the more magma is produced) and deform (the more water there is, the less viscous they are). These two properties alone emphasize that to understand the distribution of volcanism and the mechanism of plate tectonics, the water content of the mantle must be determined - Earth being a template to which all other terrestrial planets can be compared.

  19. Efficacy of sanitized ice in reducing bacterial load on fish fillet and in the water collected from the melted ice.

    Science.gov (United States)

    Feliciano, Lizanel; Lee, Jaesung; Lopes, John A; Pascall, Melvin A

    2010-05-01

    This study investigated the efficacy of sanitized ice for the reduction of bacteria in the water collected from the ice that melted during storage of whole and filleted Tilapia fish. Also, bacterial reductions on the fish fillets were investigated. The sanitized ice was prepared by freezing solutions of PRO-SAN (an organic acid formulation) and neutral electrolyzed water (NEW). For the whole fish study, the survival of the natural microflora was determined from the water of the melted ice prepared with PRO-SAN and tap water. These water samples were collected during an 8 h storage period. For the fish fillet study, samples were inoculated with Escherichia coli K12, Listeria innocua, and Pseudomonas putida then stored on crushed sanitized ice. The efficacies of these were tested by enumerating each bacterial species on the fish fillet and in the water samples at 12 and 24 h intervals for 72 h, respectively. Results showed that each bacterial population was reduced during the test. However, a bacterial reduction of fillet samples. A maximum of approximately 2 log CFU and > 3 log CFU reductions were obtained in the waters sampled after the storage of whole fish and the fillets, respectively. These reductions were significantly (P < 0.05) higher in the water from sanitized ice when compared with the water from the unsanitized melted ice. These results showed that the organic acid formulation and NEW considerably reduced the bacterial numbers in the melted ice and thus reduced the potential for cross-contamination.

  20. 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)

  1. Deducing Water Concentrations in the Parent Magma of Cumulate Clinopyroxene and Olivine: Implications for a Hydrous Parent Melt of a Primitive Deccan Lava

    Science.gov (United States)

    Seaman, S. J.

    2017-12-01

    Water concentrations of clinopyroxene megacrysts in the Powai ankaramite flow, located near Mumbai, Deccan province, India, indicate that the parent magma of the flow hosted at least 4.3 wt.% water, an unusually high water concentration for a continental flood basalt magma. The Powai flow hosts clinopyroxene and olivine phenocrysts. Chatterjee and Sheth (2015) showed that phenocrysts in the flow were part of a cumulate layer intruded by basaltic melt at 6 kb and 1230oC, so the phenocrysts record characteristics of the cumulate parent melt. Clinopyroxene phenocrysts are oscillatorily zoned in water, Mg, Fe, and Ca concentrations, and have concentric bands 100-200 microns thick of 10-20 micron diameter melt inclusions. Olivine phenocrysts host only larger isolated melt inclusions. Zones in the cpx phenocrysts where melt inclusion-rich concentric bands occur have higher concentrations of water than inclusion-free zones. Water concentrations of cpx were used to calculate water concentrations in the melt from which the crystals formed using partition coefficients of Hauri et al. (2004). Water concentrations in the parent magma were between 4.3 and 8.2 wt. % based on water concentrations in cpx. Both Mg and Fe are relatively depleted in the water- and melt inclusion-rich zones in cpx, and Ca is enriched in these zones. Oscillatory zoning in cpx may be a result of repeated growth of cpx in water- richer and water-poorer boundary layers where water lowered melt viscosity and enhanced diffusion and crystal growth rates. Water-enhanced growth rates may have resulted in capture of melt inclusions preserved in water-rich cpx zones. Melt inclusions in olivine phenocrysts preserve lower water concentrations ( 1.2 wt. %) than those indicated by water concentration in cpx phenocrysts. This disparity may be evidence of water loss from melt inclusions in olivine (Gaetani et al., 2009) or may indicate that cpx and ol crystals did not crystallize from the same parent at the same time.

  2. 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

  3. Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

    2017-12-15

    This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

  4. How do the radiative effects of springtime clouds and water vapor modulate the melt onset of Arctic sea ice?

    Science.gov (United States)

    Huang, Y.; Dong, X.; Xi, B.; Deng, Y.

    2017-12-01

    Earlier studies show that there is a strong positive correlation between the mean onset date of snow melt north of 70°N and the minimum Arctic sea ice extent (SIE) in September. Based on satellite records from 1980 to 2016, the September Arctic SIE minimum is most sensitive to the early melt onset over the Siberian Sea (73°-84°N, 90°-155°), which is defined as the area of focus (AOF) in this analysis. The day with melt onset exceeding 10% area of the AOF is marked as the initial melt date for a given year. With this definition, a strong positive correlation (r=0.59 at 99% confidence level) is found between the initial melt date over the AOF and the September SIE minimum over the Arctic. Daily anomalies of cloud and radiation properties are compared between six years with earliest initial melt dates (1990, 2012, 2007, 2003, 1991, 2016) and six years with latest initial melt dates (1996, 1984, 1983, 1982, 1987, 1992) using the NASA MERRA-2 reanalysis. Our results suggest that higher cloud water path (CWP) and precipitable water vapor (PWV) are clearly associated with early melt onset years through the period of mid-March to August. Major contrasts in CWP are found between the early and late onset years in a period of approximately 30 days prior to the onset to 30 days after the onset. As a result, the early melt onset years exhibit positive anomalies for downward longwave flux at the surface and negative anomalies for downward shortwave flux, shortwave cloud radiative effect (CRE) as well as net CRE. The negative net CRE is over-compensated by the positive longwave flux anomaly associated with elevated PWV, contributing to early melt onsets. The temporal evolution of CRE and PWV radiative effect during the entire melting season will be documented together with an analysis tracing the dynamical, mid-latitude origins of increased CWP and PWV prior to initial melt onsets.

  5. Quantifying present and future glacier melt-water contribution to runoff in a central Himalayan river basin

    Directory of Open Access Journals (Sweden)

    M. Prasch

    2013-05-01

    Full Text Available Water supply of most lowland cultures heavily depends on rain and melt water from the upstream mountains. Especially melt-water release of alpine mountain ranges is usually attributed a pivotal role for the water supply of large downstream regions. Water scarcity is assumed as consequence of glacier shrinkage and possible disappearance due to global climate change (GCC, in particular for large parts of Central and Southeast Asia. In this paper, the application and validation of a coupled modeling approach with regional climate model (RCM outputs and a process-oriented glacier and hydrological model is presented for the central Himalayan Lhasa River basin despite scarce data availability. Current and possible future contributions of ice melt to runoff along the river network are spatially explicitly shown. Its role among the other water balance components is presented. Although glaciers have retreated and will continue to retreat according to the chosen climate scenarios, water availability is and will be primarily determined by monsoon precipitation and snowmelt. Ice melt from glaciers is and will be a minor runoff component in summer monsoon-dominated Himalayan river basins.

  6. Regions of open water and melting sea ice drive new particle formation in North East Greenland.

    Science.gov (United States)

    Dall Osto, M; Geels, C; Beddows, D C S; Boertmann, D; Lange, R; Nøjgaard, J K; Harrison, Roy M; Simo, R; Skov, H; Massling, A

    2018-04-17

    Atmospheric new particle formation (NPF) and growth significantly influences the indirect aerosol-cloud effect within the polar climate system. In this work, the aerosol population is categorised via cluster analysis of aerosol number size distributions (9-915 nm, 65 bins) taken at Villum Research Station, Station Nord (VRS) in North Greenland during a 7 year record (2010-2016). Data are clustered at daily averaged resolution; in total, we classified six categories, five of which clearly describe the ultrafine aerosol population, one of which is linked to nucleation events (up to 39% during summer). Air mass trajectory analyses tie these frequent nucleation events to biogenic precursors released by open water and melting sea ice regions. NPF events in the studied regions seem not to be related to bird colonies from coastal zones. Our results show a negative correlation (r = -0.89) between NPF events and sea ice extent, suggesting the impact of ultrafine Arctic aerosols is likely to increase in the future, given the likely increased sea ice melting. Understanding the composition and the sources of Arctic aerosols requires further integrated studies with joint multi-component ocean-atmosphere observation and modelling.

  7. Effect of coolant velocity on the fragmentation of single melt drops in water

    International Nuclear Information System (INIS)

    Cunningham, M.H.; Frost, D.L.

    1997-01-01

    Flash X-ray radiography and high-speed photography are used to investigate the effect of the coolant velocity on the fine fragmentation of molten tin drops in water. A water cannot is used to accelerate the water to a constant speed of up to 30 m/s. The water is accelerated with a double piston arrangement including a foam shock absorber to eliminate the formation of a shock wave. In this way, the effect of coolant velocity on drop breakup is investigated in the absence of the strong shock wave that is present in most earlier studies. The results show that there is a transition from thermal to hydrodynamic fragmentation through an intermediate stage in which the drops initially undergo hydrodynamic fragmentation followed by the formation of a vapour bubble. For low velocities (9 m/s) this bubble collapses, fragmenting the remainder of the drop while at greater velocities (15 m/s) the drop breaks up within the bubble before it condenses. At 22 and 28 m/s there is no vapour formation and the drop fragments due to hydrodynamic effects. Quantitative analysis of the radiographs is used to determine the mass distribution of the melt during the drop fragmentation. Comparison with earlier work in which the ambient flow is preceded by a strong shock wave indicates that the transition from thermal to hydrodynamic breakup is strongly dependent on the characteristics of the pressure field experienced by the drop. (author)

  8. Holocene evolution of a drowned melt-water valley in the Danish Wadden Sea

    DEFF Research Database (Denmark)

    Pedersen, Jørn Bjarke Torp; Svinth, Steffen; Bartholdy, Jesper

    2009-01-01

    Cores from the salt marshes along the drowned melt-water valley of river Varde Å in the Danish Wadden Sea have been dated and analysed (litho- and biostratigraphically) to reconstruct the Holocene geomorphologic evolution and relative sea level history of the area. The analysed cores cover...... the total post-glacial transgression, and the reconstructed sea level curve represents the first unbroken curve of this kind from the Danish Wadden Sea, including all phases from the time where sea level first reached the Pleistocene substrate of the area. The sea level has been rising from - 12 m below...... the present level at c. 8400 cal yr BP, interrupted by two minor drops of sea level rise, and the Holocene sequence consists in most places of clay atop...

  9. High water contents in basaltic melt inclusions from Arenal volcano, Costa Rica

    Science.gov (United States)

    Wade, J. A.; Plank, T.; Hauri, E. H.; Melson, W. G.; Soto, G. J.

    2004-12-01

    Despite the importance of water to arc magma genesis, fractionation and eruption, few quantitative constraints exist on the water content of Arenal magmas. Early estimates, by electron microprobe sum deficit, suggested up to 4 wt% H2O in olivine-hosted basaltic andesite melt inclusions (MI) from pre-historic ET-6 tephra (Melson, 1982), and up to 7 wt% H2O in plagioclase and orthopyroxene-hosted dacitic MI from 1968 lapilli (Anderson, 1979). These high water contents are consistent with abundant hornblende phenocrysts in Arenal volcanics, but inconsistent with geochemical tracers such as 10Be and Ba/La that suggest a low flux of recycled material (and presumably water) from the subduction zone. In order to test these ideas, and provide the first direct measurements of water in mafic Arenal magmas, we have studied olivine-hosted MI from the prehistoric (900 yBP; Soto et al., 1998) ET3 tephra layer. MI range from andesitic (> 58% SiO2) to basaltic compositions ( 4 wt%) found here for Arenal basaltic MI support the semi-quantitative data from earlier studies, but are somewhat unexpected given predictions from slab tracers. Arenal water contents (4%) approach those of the 1995 eruption of Cerro Negro in Nicaragua (4-5 wt% in basaltic MI; Roggensack et al., 1997), despite the fact that the latter has Ba/La of > 100, while Arenal has Ba/La Journal of Geology; Melson, William G. (1982) Boletin de Volcanologia; Roggensack et al. (1997) Science; Soto et al. (1998) OSIVAM; Williams-Jones et al. (2001) Journal of Volc. and Geoth. Res.

  10. Transformation of irregular shaped silver nanostructures into nanoparticles by under water pulsed laser melting

    Science.gov (United States)

    Yadavali, S.; Sandireddy, V. P.; Kalyanaraman, R.

    2016-05-01

    The ability to easily manufacture nanostructures with a desirable attribute, such as well-defined size and shape, especially from any given initial shapes or sizes of the material, will be helpful towards accelerating the use of nanomaterials in various applications. In this work we report the transformation of discontinuous irregular nanostructures (DIN) of silver metal by rapid heating under a bulk fluid layer. Ag films were changed into DIN by dewetting in air and subsequently heated by nanosecond laser pulses under water. Our findings show that the DIN first ripens into elongated structures and then breaks up into nanoparticles. From the dependence of this behavior on laser fluence we found that under water irradiation reduced the rate of ripening and also decreased the characteristic break-up length scale of the elongated structures. This latter result was qualitatively interpreted as arising from a Rayleigh-Plateau instability modified to yield significantly smaller length scales than the classical process due to pressure gradients arising from the rapid evaporation of water during laser melting. These results demonstrate that it is possible to fabricate a dense collection of monomodally sized Ag nanoparticles with significantly enhanced plasmonic quality starting from the irregular shaped materials. This can be beneficial towards transforming discontinuous Ag films into nanostructures with useful plasmonic properties, that are relevant for biosensing applications.

  11. Water solubility in monzogranite melts: experimental and calculated water contents at 6 kbar

    OpenAIRE

    García Moreno, Olga; Castro Dorado, Antonio; Corretgé, Luis Guillermo

    2002-01-01

    Several piston-cylinder crystallisation experiments have been performed with a synthetic monzogranitic glass with different initial water contents at 6 kbar. Comparison with calculated water contents shows: 1) some differences of the order of 10% of XH2Q; 2) "non-linear" behaviour in XH2C/T curves; and 3) similar pattern in the XH2JT curves in both measured and calculated data. Resumen Se han realizado varios experimentos de cristalización en aparatos "piston-cylinder" a 6 kbar, u...

  12. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    International Nuclear Information System (INIS)

    Tran, Chi Thanh

    2009-09-01

    indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR

  13. Ex-vessel melt-coolant interactions in deep water pool: Studies and accident management for Swedish BWRs

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Chu, C.C.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

    1993-01-01

    In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL analyses

  14. Ex-vessel melt-coolant interactions in deep water pool: studies and accident management for Swedish BWRs

    International Nuclear Information System (INIS)

    Chu, C.C.; Sienicki, J.J.; Spencer, B.W.; Frid, W.; Loewenhielm, G.

    1995-01-01

    In Swedish BWRs having an annular suppression pool, the lower drywell beneath the reactor vessel is flooded with water to mitigate against the effects of melt release into the drywell during a severe accident. The THIRMAL-1 code has been used to analyze the effectiveness of the water pool to protect lower drywell penetrations by fragmenting and quenching the melt as it relocates downward through the water. Experiments have also been performed to investigate the benefits of adding surfactants to the water to reduce the likelihood of fine-scale debris formation from steam explosions. This paper presents an overview of the accident management approach and surfactant investigations together with results from the THIRMAL-1 analyses. A description of the modeling incorporated in THIRMAL-1 is also provided. (orig.)

  15. Hydro-Chemical Characterization of Melt Waters of Ponkar Glacier, Manang, Nepal

    Science.gov (United States)

    Shrestha, R.; Sandeep, S.

    2016-12-01

    The study was carried out in Ponkar Glacier, representing Himalayan glacier of Nepal. The study aims in determining the hydrochemistry of the glacier melt water. The sampling sites included moraine dammed, Ponkar Lake at 4120 m a.s.l to the downstream glaciated river at 3580 m a.s.l. The water samples were collected from the seven different sites according to the geological features. On site measurements of the parameters like pH, temperature, electrical conductivity was done by digital multi-probes. The samples were brought to the laboratory and the parameters were analyzed according to the standard guidelines and protocols. The glacier meltwater was slightly basic with pH 7.44 (±0.32). The water was slightly hard 36.43 (±9.15) mg CaCO3/L and the electrical conductivity was found to be 47.14 (±11.18) µS/cm. The concentration of anion was in the order of HCO3 - > Cl- > SO42- > NO3- > PO43-. Calcium carbonate weathering was found out to be the major source of dissolved ions in the region. The parameters like chloride, total silica and iron were found to be 55.71 (±32.03) mg/L, 1.13 (±0.76) mg/L and 1.1 (±0.97) mg/L which is in the significant range. Whereas the concentration of manganese (<0.05 mg/L) and zinc (<0.02 mg/L) were not in detectable levels in few stations. The results of this study can be helpful in preliminary assessment of hydrochemistry and its linkage with climate change impacts in this region.

  16. Methane excess in Arctic surface water-triggered by sea ice formation and melting.

    Science.gov (United States)

    Damm, E; Rudels, B; Schauer, U; Mau, S; Dieckmann, G

    2015-11-10

    Arctic amplification of global warming has led to increased summer sea ice retreat, which influences gas exchange between the Arctic Ocean and the atmosphere where sea ice previously acted as a physical barrier. Indeed, recently observed enhanced atmospheric methane concentrations in Arctic regions with fractional sea-ice cover point to unexpected feedbacks in cycling of methane. We report on methane excess in sea ice-influenced water masses in the interior Arctic Ocean and provide evidence that sea ice is a potential source. We show that methane release from sea ice into the ocean occurs via brine drainage during freezing and melting i.e. in winter and spring. In summer under a fractional sea ice cover, reduced turbulence restricts gas transfer, then seawater acts as buffer in which methane remains entrained. However, in autumn and winter surface convection initiates pronounced efflux of methane from the ice covered ocean to the atmosphere. Our results demonstrate that sea ice-sourced methane cycles seasonally between sea ice, sea-ice-influenced seawater and the atmosphere, while the deeper ocean remains decoupled. Freshening due to summer sea ice retreat will enhance this decoupling, which restricts the capacity of the deeper Arctic Ocean to act as a sink for this greenhouse gas.

  17. Water in melt inclusions from phenocrysts of dacite pumice of the Vetrovoy Isthmus (Iturup Island, Southern Kuriles)

    Science.gov (United States)

    Kotov, A. A.; Smirnov, S. Z.; Maksimovich, I. A.; Plechov, P. Yu; Chertkova, N. V.; Befus, A. I.

    2017-12-01

    This work is devoted to the study of one of the largest caldera eruptions of the Kurile-Kamchatka island-arc system that occurred on the island of Iturup. The object of investigation of this work are phenocrysts of quartz and plagioclase from dacite pumice of the Isthmus of the Isthmus, which is located on the island of Iturup. The purpose of this work is to determine the water content in the melts that participated in the caldera eruption of the Vetrovoy Isthmus and the patterns of their changes during the crystallization of magma. In the course of the work, the following were carried out: 1) adaptation and calibration of the Raman spectroscopy method for determining water in rhyolite melt’s inclusions glasses in quartz and plagioclase from pumice stone; 2) determination of composition and estimation of water content in melt inclusions in quartz and plagioclase according to x-ray spectral analysis; 3) establishment of the regularities of the change in the water content during the evolution of the magmatic melt; 4) evaluation of fluid pressure by comparison with experimental data

  18. Hydro-chemical Characterization of Glacier Melt Water of Ponkar Glacier, Manang, Nepal.

    Science.gov (United States)

    Shrestha, R.; Sandeep, S.

    2017-12-01

    The study was carried out in Ponkar Glacier, representing Himalayan glacier of Nepal. The study aims in determining the physical-chemical properties of the glacier melt water. The sampling sites included moraine dammed, Ponkar Lake at 4100 m a.s.l to the downstream glaciated stream at 3580 m a.s.l. The water samples were collected from the seven different sites. Temperature was recorded by digital multi-thermometer on site. The samples were brought to the laboratory and the parameters were analyzed according to the APHA, AWWA and WEF standards. The glacier meltwater was slightly basic with pH 7.44 (±0.307). The meltwater was found to be in the range 30-60 which implies the water is moderately soft resulting value of concentration 36.429±8.664 mg CaCO3 L-1 and the electrical conductivity was found to be 47.14 (±11.18) µS/cm. The concentration of anion was in the order of HCO3 - > Cl- > SO42- > NO3- > TP-PO43- with the concentration 194.286±40.677, 55.707±30.265, 11.533±1.132 mgL-1, 1.00±0.7 mgL-1 and 0.514±0.32 mgL-1 respectively. Calcium carbonate weathering was found out to be the major source of dissolved ions in the region. The heavy metals were found in the order Al>Fe>Mn>Zn with concentration 1.34±0.648, 1.103±0.917, 0.08±0.028 and 0.023±0.004 mgL-1 respectively. The concentration of iron, manganese and zinc in some sites were below the detection limit. These results represent baseline data for the physical-chemical properties of the glacier meltwater

  19. Vulnerability of Southeast Greenland Glaciers to Warm Atlantic Water From Operation IceBridge and Ocean Melting Greenland Data

    Science.gov (United States)

    Millan, R.; Rignot, E.; Mouginot, J.; Wood, M.; Bjørk, A. A.; Morlighem, M.

    2018-03-01

    We employ National Aeronautics and Space Administration (NASA)'s Operation IceBridge high-resolution airborne gravity from 2016, NASA's Ocean Melting Greenland bathymetry from 2015, ice thickness from Operation IceBridge from 2010 to 2015, and BedMachine v3 to analyze 20 major southeast Greenland glaciers. The results reveal glacial fjords several hundreds of meters deeper than previously thought; the full extent of the marine-based portions of the glaciers; deep troughs enabling warm, salty Atlantic Water (AW) to reach the glacier fronts and melt them from below; and few shallow sills that limit the access of AW. The new oceanographic and topographic data help to fully resolve the complex pattern of historical ice front positions from the 1930s to 2017: glaciers exposed to AW and resting on retrograde beds have retreated rapidly, while glaciers perched on shallow sills or standing in colder waters or with major sills in the fjords have remained stable.

  20. Flocculation alters the distribution and flux of melt-water supplied sediments and nutrients in the Arctic

    DEFF Research Database (Denmark)

    Markussen, Thor Nygaard; Andersen, Thorbjørn Joest; Ernstsen, Verner Brandbyge

    In the Arctic, thawing permafrost and increased melting of glaciers are important drivers for changes in fine-grained sediment supply and biogeochemical fluxes from land to sea. Flocculation of particles is a controlling factor for the magnitude of fluxes and deposition rates in the marine...... environment but comparatively little is known about the flocculation processes in the Arctic. We investigated flocculation dynamics from a melt-water river in the inner Disko Fjord, West Greenland. A novel, laser-illuminated camera system significantly improved the particle size measurement capabilities...... and settling tubes were sampled to enable sub-sampling of different floc size fractions. Flocculation was observed during periods with low turbulent shear and also at the front of the fresh water plume resulting in significant volumes of large sized flocs at depth below the plume. The floc sizes and volumes...

  1. 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)

  2. 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

  3. 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

  4. In situ study at high pressure and temperature of the environment of water in hydrous Na and Ca aluminosilicate melts and coexisting aqueous fluids

    Science.gov (United States)

    Le Losq, Charles; Dalou, Célia; Mysen, Bjorn O.

    2017-07-01

    The bonding and speciation of water dissolved in Na silicate and Na and Ca aluminosilicate melts were inferred from in situ Raman spectroscopy of the samples, in hydrothermal diamond anvil cells, while at crustal temperature and pressure conditions. Raman data were also acquired on Na silicate and Na and Ca aluminosilicate glasses, quenched from hydrous melts equilibrated at high temperature and pressure in a piston cylinder apparatus. In the hydrous melts, temperature strongly influences O-H stretching ν(O-H) signals, reflecting its control on the bonding of protons between different molecular complexes. Pressure and melt composition effects are much smaller and difficult to discriminate with the present data. However, the chemical composition of the melt + fluid system influences the differences between the ν(O-H) signals from the melts and the fluids and, hence, between their hydrogen partition functions. Quenching modifies the O-H stretching signals: strong hydrogen bonds form in the glasses below the glass transition temperature Tg, and this phenomenon depends on glass composition. Therefore, glasses do not necessarily record the O-H stretching signal shape in melts near Tg. The melt hydrogen partition function thus cannot be assessed with certainty using O-H stretching vibration data from glasses. From the present results, the ratio of the hydrogen partition functions of hydrous silicate melts and aqueous fluids mostly depends on temperature and the bulk melt + fluid system chemical composition. This implies that the fractionation of hydrogen isotopes between magmas and aqueous fluids in water-saturated magmatic systems with differences in temperature and bulk chemical composition will be different.

  5. 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.)

  6. Modeling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R. [Royal Institute of Technology, Stockholm (Sweden)] [and others

    1995-09-01

    The objective of the paper is to study heat and mass transfer processes related to core melt discharge from a reactor vessel is a severe light water reactor accident. The phenomenology of the issue includes (1) melt convection in and heat transfer from the melt pool in contact with the vessel lower head wall; (2) fluid dynamics and heat transfer of the melt flow in the growing discharge hole; and (3) multi-dimensional heat conduction in the ablating lower head wall. A program of model development, validation and application is underway (i) to analyse the dominant physical mechanisms determining characteristics of the lower head ablation process; (ii) to develop and validate efficient analytic/computational methods for estimating heat and mass transfer under phase-change conditions in irregular moving-boundary domains; and (iii) to investigate numerically the melt discharge phenomena in a reactor-scale situation, and, in particular, the sensitivity of the melt discharge transient to structural differences and various in-vessel melt progression scenarios. The paper presents recent results of the analysis and model development work supporting the simulant melt-structure interaction experiments.

  7. Melt spreading code assessment, modifications, and initial application to the EPR core catcher design

    International Nuclear Information System (INIS)

    Farmer, M.T.; Basu, S.

    2009-01-01

    The Evolutionary Power Reactor (EPR) is a 1,600-MWe Pressurized Water Reactor (PWR) that is undergoing a design certification review by the U.S. Nuclear Regulatory Commission (NRC). The EPR severe accident design philosophy is predicated upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external flooding. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: 1) an external core melt retention system to temporarily hold core melt released from the vessel; 2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; 3) a melt plug that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, 4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and non-uniform spreading. The NRC is using MELTSPREAD to evaluate melt spreading in the EPR design. The development of MELTSPREAD ceased in the early 1990's, and so the code was first assessed against the more contemporary spreading database and code modifications, as warranted, were carried out before performing confirmatory plant calculations. This paper provides principle findings from the MELTSPREAD assessment activities and resulting code modifications, and also summarizes the results of initial scoping calculations for the EPR plant design and preliminary plant analyses, along with the plan for performing the final set of plant calculations including sensitivity studies

  8. In-vessel phenomena -- CORA

    International Nuclear Information System (INIS)

    Ott, L.J.; Rij, W.I. van.

    1991-01-01

    Experiment-specific models have been employed since 1986 by Oak Ridge National Laboratory (ORNL) severe accident analysis programs for the purpose of boiling water reactor experimental planning and optimum interpretation of experimental results. The large integral tests performed to date, which start from an initial undamaged core state, have involved significantly different-from-prototypic boundary and experimental conditions because of either normal facility limitations or specific experimental constraints. These experiments (ACRR: DF-4, NRU: FLHT-6, and CORA) were designed to obtain specific phenomenological information such as the degradation and interaction of prototypic components and the effects on melt progression of control-blade materials and channel boxes. Applications of ORNL models specific to the KfK CORA-16 and CORA-17 experiments are discussed and significant findings from the experimental analyses such as the following are presented: applicability of available Zircaloy oxidation kinetics correlations; influence of cladding strain on Zircaloy oxidation; influence of spacer grids on the structural heatup; and the impact of treating the gaseous coolant as a gray interacting medium. The experiment-specific models supplement and support the systems-level accident analysis codes. They allow the analyst to accurately quantify the observed experimental phenomena and to compensate for the effect of known uncertainties. They provide a basis for the efficient development of new models for phenomena that are currently not modeled (such as material interactions). They can provide validated phenomenological models (from the results of the experiments) as candidates for incorporation in the systems-level ''whole-core'' codes

  9. Diazotroph Diversity in the Sea Ice, Melt Ponds, and Surface Waters of the Eurasian Basin of the Central Arctic Ocean.

    Science.gov (United States)

    Fernández-Méndez, Mar; Turk-Kubo, Kendra A; Buttigieg, Pier L; Rapp, Josephine Z; Krumpen, Thomas; Zehr, Jonathan P; Boetius, Antje

    2016-01-01

    The Eurasian basin of the Central Arctic Ocean is nitrogen limited, but little is known about the presence and role of nitrogen-fixing bacteria. Recent studies have indicated the occurrence of diazotrophs in Arctic coastal waters potentially of riverine origin. Here, we investigated the presence of diazotrophs in ice and surface waters of the Central Arctic Ocean in the summer of 2012. We identified diverse communities of putative diazotrophs through targeted analysis of the nifH gene, which encodes the iron protein of the nitrogenase enzyme. We amplified 529 nifH sequences from 26 samples of Arctic melt ponds, sea ice and surface waters. These sequences resolved into 43 clusters at 92% amino acid sequence identity, most of which were non-cyanobacterial phylotypes from sea ice and water samples. One cyanobacterial phylotype related to Nodularia sp. was retrieved from sea ice, suggesting that this important functional group is rare in the Central Arctic Ocean. The diazotrophic community in sea-ice environments appear distinct from other cold-adapted diazotrophic communities, such as those present in the coastal Canadian Arctic, the Arctic tundra and glacial Antarctic lakes. Molecular fingerprinting of nifH and the intergenic spacer region of the rRNA operon revealed differences between the communities from river-influenced Laptev Sea waters and those from ice-related environments pointing toward a marine origin for sea-ice diazotrophs. Our results provide the first record of diazotrophs in the Central Arctic and suggest that microbial nitrogen fixation may occur north of 77°N. To assess the significance of nitrogen fixation for the nitrogen budget of the Arctic Ocean and to identify the active nitrogen fixers, further biogeochemical and molecular biological studies are needed.

  10. Diazotroph diversity in the sea ice, melt ponds and surface waters of the Eurasian Basin of the Central Arctic Ocean

    Directory of Open Access Journals (Sweden)

    Mar Fernández-Méndez

    2016-11-01

    Full Text Available The Eurasian basin of the Central Arctic Ocean is nitrogen limited, but little is known about the presence and role of nitrogen-fixing bacteria. Recent studies have indicated the occurrence of diazotrophs in Arctic coastal waters potentially of riverine origin. Here, we investigated the presence of diazotrophs in ice and surface waters of the Central Arctic Ocean in the summer of 2012. We identified diverse communities of putative diazotrophs through targeted analysis of the nifH gene, which encodes the iron protein of the nitrogenase enzyme. We amplified 529 nifH sequences from 26 samples of Arctic melt ponds, sea ice and surface waters. These sequences resolved into 43 clusters at 92% amino acid sequence identity, most of which were non-cyanobacterial phylotypes from sea ice and water samples. One cyanobacterial phylotype related to Nodularia sp. was retrieved from sea ice, suggesting that this important functional group is rare in the Central Arctic Ocean. The diazotrophic community in sea-ice environments appear distinct from other cold-adapted diazotrophic communities, such as those present in the coastal Canadian Arctic, the Arctic tundra and glacial Antarctic lakes. Molecular fingerprinting of nifH and the intergenic spacer region of the rRNA operon revealed differences between the communities from river-influenced Laptev Sea waters and those from ice-related environments pointing towards a marine origin for sea-ice diazotrophs. Our results provide the first record of diazotrophs in the Central Arctic and suggest that microbial nitrogen fixation may occur north of 77ºN. To assess the significance of nitrogen fixation for the nitrogen budget of the Arctic Ocean and to identify the active nitrogen fixers, further biogeochemical and molecular biological studies are needed.

  11. A coupled melt-freeze temperature index approach in a one-layer model to predict bulk volumetric liquid water content dynamics in snow

    Science.gov (United States)

    Avanzi, Francesco; Yamaguchi, Satoru; Hirashima, Hiroyuki; De Michele, Carlo

    2016-04-01

    Liquid water in snow rules runoff dynamics and wet snow avalanches release. Moreover, it affects snow viscosity and snow albedo. As a result, measuring and modeling liquid water dynamics in snow have important implications for many scientific applications. However, measurements are usually challenging, while modeling is difficult due to an overlap of mechanical, thermal and hydraulic processes. Here, we evaluate the use of a simple one-layer one-dimensional model to predict hourly time-series of bulk volumetric liquid water content in seasonal snow. The model considers both a simple temperature-index approach (melt only) and a coupled melt-freeze temperature-index approach that is able to reconstruct melt-freeze dynamics. Performance of this approach is evaluated at three sites in Japan. These sites (Nagaoka, Shinjo and Sapporo) present multi-year time-series of snow and meteorological data, vertical profiles of snow physical properties and snow melt lysimeters data. These data-sets are an interesting opportunity to test this application in different climatic conditions, as sites span a wide latitudinal range and are subjected to different snow conditions during the season. When melt-freeze dynamics are included in the model, results show that median absolute differences between observations and predictions of bulk volumetric liquid water content are consistently lower than 1 vol%. Moreover, the model is able to predict an observed dry condition of the snowpack in 80% of observed cases at a non-calibration site, where parameters from calibration sites are transferred. Overall, the analysis show that a coupled melt-freeze temperature-index approach may be a valid solution to predict average wetness conditions of a snow cover at local scale.

  12. Evaluation of In-Vessel Corium Retention under a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae-Joon; Kang, Kyung-Ho; Ha, Kwang-Soon; Kim, Jong-Tae; Koo, Kil-Mo; Cho, Young-Ro; Hong, Seong-Wan; Kim, Sang-Baik; Kim, Hee-Dong

    2008-02-15

    The current study on In-Vessel corium Retention and its application activities to the actual nuclear power plant have been reviewed and discussed in this study. Severe accident sequence which determines an initial condition of the IVR has been evaluated and late phase melt progression, heat transfer on the outer reactor vessel, and in-vessel corium cooling mechanism have been estimated in detail. During the high pressure sequence of the reactor coolant system, a natural circulation flow of the hot steam leads to a failure of the pressurizer surge line before the reactor vessel failure, which leads to a rapid decrease of the reactor coolant system pressure. The results of RASPLAV/MASCA study by OECD/NEA have shown that a melt stratification has occurred in the lower plenum of the reactor vessel. In particular, laver inversion has occurred, which is that a high density of the metal melt moves to the lower part of the oxidic melt layer. A method of heat transfer enhancement on the outer reactor vessel is an optimal design of the reactor vessel insulation for an increase of the natural circulation flow between the outer reactor vessel and the its insulation, and an increase of the critical Heat flux on the outer reactor vessel by using various method, such as Nono fluid, coated reactor vessel, and so on. An increase method of the in-vessel melt cooling is a development of the In-vessel core catcher and a decrease of focusing effect in the metal layer.

  13. The effects of sub-ice-shelf melting on dense shelf water formation and export in idealized simulations of Antarctic margins

    Science.gov (United States)

    Marques, Gustavo; Stern, Alon; Harrison, Matthew; Sergienko, Olga; Hallberg, Robert

    2017-04-01

    Dense shelf water (DSW) is formed in coastal polynyas around Antarctica as a result of intense cooling and brine rejection. A fraction of this water reaches ice shelves cavities and is modified due to interactions with sub-ice-shelf melt water. This modified water mass contributes to the formation of Antarctic Bottom Water, and consequently, influences the large-scale ocean circulation. Here, we investigate the role of sub-ice-shelf melting in the formation and export of DSW using idealized simulations with an isopycnal ocean model (MOM6) coupled with a sea ice model (SIS2) and a thermodynamic active ice shelf. A set of experiments is conducted with variable horizontal grid resolutions (0.5, 1.0 and 2.0 km), ice shelf geometries and atmospheric forcing. In all simulations DSW is spontaneously formed in coastal polynyas due to the combined effect of the imposed atmospheric forcing and the ocean state. Our results show that sub-ice-shelf melting can significantly change the rate of dense shelf water outflows, highlighting the importance of this process to correctly represent bottom water formation.

  14. Temperatures and Melt Water Contents at the Onset of Phenocryst Growth in Quaternary Nepheline-Normative Basalts Erupted along the Tepic-Zacoalco Rift in Western Mexico

    Science.gov (United States)

    Mesa, J.; Lange, R. A.; Pu, X.

    2017-12-01

    Nepheline-normative, high-Mg basalts erupted from the western Mexican arc, along the Tepic-Zacoalco rift (TZR), have a trace-element signature consistent with an asthenosphere source, whereas calc-alkaline basalts erupted from the central Mexican arc in the Michoacan-Guanajuato volcanic field (MGVF) have a trace-element signature consistent with a mantle source strongly affected by subduction fluids. In this study, olivine-melt thermometry and plagioclase-liquid hygrometry are used to constrain the temperature and melt water content of the alkaline TZR basalts. The presence of diffusion-limited growth textures in olivine and plagioclase phenocrysts provide preliminary evidence of rapid growth during ascent. For each basalt sample, a histogram of all analyzed olivines in each sample allows the most Fo-rich composition to be identified, which matches the calculated composition at the liquidus via MELTS (Ghiorso & Sack, 1995; Asimow & Ghiorso, 1998) at fO2 values of QFM +2. Therefore a newly developed olivine-melt thermometer, based on DNiol/liq (Pu et al., 2017) was used to calculate temperature at the onset of olivine crystallization during ascent. Temperatures range from 1076-1247°C, whereas those calculated using an olivine-melt thermometer based on DMgol/liq range from 1141-1236 °C. Olivine-melt thermometers based on DMgol/liq are sensitive to melt H2O content, therefore ΔT = TMg - TNi (≤ 82 degrees) may be used as a qualitative indicator of melt H2O (≤ 2.6 wt% H2O; Pu et al., 2017). When temperatures from the Ni-thermometer are applied to the most calcic plagioclase in each sample (Waters & Lange, 2015), calculated melt H2O contents range from 1.3-1.9 (± 0.4) wt%. These values are significantly lower than those obtained from high-Mg calc-alkaline basalts from the MGVF using similar methods (1.9-5.0 wt%; Pu et al., 2017), consistent with a reduced involvement of slab-derived fluids in the origin of the alkaline TZR basalts from western Mexico.

  15. The great melting pot. Common sole population connectivity assessed by otolith and water fingerprints.

    Science.gov (United States)

    Morat, Fabien; Letourneur, Yves; Dierking, Jan; Pécheyran, Christophe; Bareille, Gilles; Blamart, Dominique; Harmelin-Vivien, Mireille

    2014-01-01

    Quantifying the scale and importance of individual dispersion between populations and life stages is a key challenge in marine ecology. The common sole (Solea solea), an important commercial flatfish in the North Sea, Atlantic Ocean and the Mediterranean Sea, has a marine pelagic larval stage, a benthic juvenile stage in coastal nurseries (lagoons, estuaries or shallow marine areas) and a benthic adult stage in deeper marine waters on the continental shelf. To date, the ecological connectivity among these life stages has been little assessed in the Mediterranean. Here, such an assessment is provided for the first time for the Gulf of Lions, NW Mediterranean, based on a dataset on otolith microchemistry and stable isotopic composition as indicators of the water masses inhabited by individual fish. Specifically, otolith Ba/Ca and Sr/Ca profiles, and δ(13)C and δ(18)O values of adults collected in four areas of the Gulf of Lions were compared with those of young-of-the-year collected in different coastal nurseries. Results showed that a high proportion of adults (>46%) were influenced by river inputs during their larval stage. Furthermore Sr/Ca ratios and the otolith length at one year of age revealed that most adults (∼70%) spent their juvenile stage in nurseries with high salinity, whereas the remainder used brackish environments. In total, data were consistent with the use of six nursery types, three with high salinity (marine areas and two types of highly saline lagoons) and three brackish (coastal areas near river mouths, and two types of brackish environments), all of which contributed to the replenishment of adult populations. These finding implicated panmixia in sole population in the Gulf of Lions and claimed for a habitat integrated management of fisheries.

  16. Investigation on the hot melting temperature field simulation of HDPE water supply pipeline in gymnasium pool

    Science.gov (United States)

    Cai, Zhiqiang; Dai, Hongbin; Fu, Xibin

    2018-06-01

    In view of the special needs of the water supply and drainage system of swimming pool in gymnasium, the correlation of high density polyethylene (HDPE) pipe and the temperature field distribution during welding was investigated. It showed that the temperature field distribution has significant influence on the quality of welding. Moreover, the mechanical properties of the welded joint were analyzed by the bending test of the weld joint, and the micro-structure of the welded joint was evaluated by scanning electron microscope (SEM). The one-dimensional unsteady heat transfer model of polyethylene pipe welding joints was established by MARC. The temperature field distribution during welding process was simulated, and the temperature field changes during welding were also detected and compared by the thermo-couple temperature automatic acquisition system. Results indicated that the temperature of the end surface of the pipe does not reach the maximum value, when it is at the end of welding heating. Instead, it reaches the maximum value at 300 sand latent heat occurs during the welding process. It concludes that the weld quality is the highest when the welding pressure is 0.2 MPa, and the heating temperature of HDPE heat fusion welding is in the range of 210 °C-230 °C.

  17. FTIR microspectroscopy and SIMS study of water-poor cordierite from El Hoyazo, Spain: Application to mineral and melt devolatilization

    Science.gov (United States)

    Della Ventura, Giancarlo; Bellatreccia, Fabio; Cesare, Bernardo; Harley, Simon; Piccinini, Massimo

    2009-12-01

    This paper reports the microchemical and microspectroscopic FTIR study of cordierite from a partially melted graphite-bearing granulitic enclave within the dacitic lava dome of El Hoyazo (SE Spain). Optically transparent single-crystals, hand picked from the rock, were oriented using X-ray diffraction and studied by Fourier-transform infrared (FTIR). Single-crystal FTIR spectroscopy shows that the examined cordierite is CO 2-rich and almost H 2O-free. Two weak and sharp peaks are observed at 3708 and 3595 cm - 1 , respectively, which are strongly polarised for E // a. These peaks are assigned to combination modes of CO 2. Very weak bands due to H 2O molecules oriented with the H…H vector // c (type I water) are occasionally observed in certain zones of the grains, associated with absorptions due to hydrated inclusions of alteration products. The very intense bands observed in the 2600-2000 cm - 1 region are assigned to CO 2 molecules oriented // a; the spectra also show the presence of 13C and 18O, and weak amounts of CO in the sample. Microspectrometric mapping shows that the distribution of C is relatively homogeneous, whereas that of H 2O is complicated by a very broad absorption extending from 3700 to 3100 cm - 1 . High-resolution FTIR imaging, done using a focal-plane array of detectors, shows that this broad absorption is associated with microfractures. SIMS analyses give an average concentration of H 2O = 0.033 ± 0.007 wt.% and CO 2 = 0.21 ± 0.07 wt.%. On the basis of these data, molar absorption coefficients can be calibrated for CO 2: ɛiCO2 (integrated) = 11,000 ± 4000 l/(mol cm - 2 ) and ɛlCO2 (linear) = 800 ± 250 l/(mol cm - 1 ). Due to the extremely low amount of H 2O and its inhomogeneous distribution, calibration of absorption ɛH2O coefficients is unreliable. The very low H 2O contents in the El Hoyazo cordierite indicate continued mineral-melt volatile exchange during decompression from ˜ 5 kbar to significantly shallower levels.

  18. Insights into mercury deposition and spatiotemporal variation in the glacier and melt water from the central Tibetan Plateau.

    Science.gov (United States)

    Paudyal, Rukumesh; Kang, Shichang; Huang, Jie; Tripathee, Lekhendra; Zhang, Qianggong; Li, Xiaofei; Guo, Junming; Sun, Shiwei; He, Xiaobo; Sillanpää, Mika

    2017-12-01

    Long-term monitoring of global pollutant such as Mercury (Hg) in the cryosphere is very essential for understanding its bio-geochemical cycling and impacts in the pristine environment with limited emission sources. Therefore, from May 2015 to Oct 2015, surface snow and snow-pits from Xiao Dongkemadi Glacier and glacier melt water were sampled along an elevation transect from 5410 to 5678m a.s.l. in the central Tibetan Plateau (TP). The concentration of Hg in surface snow was observed to be higher than that from other parts of the TP. Unlike the southern parts of the TP, no clear altitudinal variation was observed in the central TP. The peak Total Hg (Hg T ) concentration over the vertical profile on the snow pits corresponded with a distinct yellowish-brown dust layer supporting the fact that most of the Hg was associated with particulate matter. It was observed that only 34% of Hg in snow was lost when the surface snow was exposed to sunlight indicating that the surface snow is less influenced by the post-depositional process. Significant diurnal variation of Hg T concentration was observed in the river water, with highest concentration observed at 7pm when the discharge was highest and lowest concentration during 7-8am when the discharge was lowest. Such results suggest that the rate of discharge was influential in the concentration of Hg T in the glacier fed rivers of the TP. The estimated export of Hg T from Dongkemadi river basin is 747.43gyr -1 , which is quite high compared to other glaciers in the TP. Therefore, the export of global contaminant Hg might play enhanced role in the Alpine regions as these glaciers are retreating at an alarming rate under global warming which may have adverse impact on the ecosystem and the human health of the region. Copyright © 2017 Elsevier B.V. All rights reserved.

  19. Melting point of yttria

    International Nuclear Information System (INIS)

    Skaggs, S.R.

    1977-06-01

    Fourteen samples of 99.999 percent Y 2 O 3 were melted near the focus of a 250-W CO 2 laser. The average value of the observed melting point along the solid-liquid interface was 2462 +- 19 0 C. Several of these same samples were then melted in ultrahigh-purity oxygen, nitrogen, helium, or argon and in water vapor. No change in the observed temperature was detected, with the exception of a 20 0 C increase in temperature from air to helium gas. Post test examination of the sample characteristics, clarity, sphericity, and density is presented, along with composition. It is suggested that yttria is superior to alumina as a secondary melting-point standard

  20. 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)

  1. 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)

  2. Investigation of catalytic oxidation of diamond by water vapor and carbon dioxide in the presence of alkali melts of some rare earth oxides

    International Nuclear Information System (INIS)

    Kulakova, I.I.; Rudenko, A.P.; Sulejmenova, A.S.; Tolstopyatova, A.A.

    1978-01-01

    The results of an investigation of the catalytic oxydation of diamond by carbon dioxide and water vapors at 906 deg C in the presence of melts of some rare earth oxides in potassium hydroxide are given. The ion La 3+ was shown to possess the most catalytic activity. The earlier proposed mechanisms of the diamond oxidation by CO 2 and H 2 O were corroborated. The ions of rare earth elements were found to accelerate the different stages of the process

  3. Application of carrier and plasticizer to improve the dissolution and bioavailability of poorly water-soluble baicalein by hot melt extrusion.

    Science.gov (United States)

    Zhang, Yilan; Luo, Rui; Chen, Yi; Ke, Xue; Hu, Danrong; Han, Miaomiao

    2014-06-01

    The objective of this study was to develop a suitable formulation for baicalein (a poorly water-soluble drug exhibiting high melting point) to prepare solid dispersions using hot melt extrusion (HME). Proper carriers and plasticizers were selected by calculating the Hansen solubility parameters, evaluating melting processing condition, and measuring the solubility of obtained melts. The characteristic of solid dispersions prepared by HME was evaluated. The dissolution performance of the extrudates was compared to the pure drug and the physical mixtures. Physicochemical properties of the extrudates were characterized by differential scanning calorimetry (DSC), powder X-ray diffraction (PXRD), and Fourier transform infrared spectroscopy (FTIR). Relative bioavailability after oral administration in beagle dogs was assessed. As a result, Kollidon VA64 and Eudragit EPO were selected as two carriers; Cremophor RH was used as the plasticizer. The dissolution of all the extrudates was significantly improved. DSC and PXRD results suggested that baicalein in the extrudates was amorphous. FTIR spectroscopy revealed the interaction between drug and polymers. After oral administration, the relative bioavailability of solid dispersions with VA64 and EPO was comparative, about 2.4- and 2.9-fold greater compared to the pure drug, respectively.

  4. Experiments on performance of the multi-layered in-vessel core catcher

    International Nuclear Information System (INIS)

    Kang, K.H.; Kim, S.B.; Park, R.J.; Cheung, F.B.; Suh, K.Y.; Rempe, J.L.

    2004-01-01

    LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with a uniform gap of 10 mm. Until now, two types of the in-vessel core catcher were used in this study. The first one is a single layered in-vessel core catcher without an internal coating of the LAVA-GAP-2 test, and the other one is a two layered in-vessel core catcher with a 0.5 mm-thick ZrO 2 internal coating of the LAVA-GAP-3 test. Current LAVA-GAP experimental results indicate that an internally coated in-vessel core catcher has better thermal performance compared with an uncoated in-vessel core catcher. Metallurgical inspections on the test specimens of the LAVA-GAP-3 test have been performed to examine the performance of the coating material and the base carbon steel. Although the base carbon steel had experienced a severe thermal attack to the extent that the microstructures were changed and re-crystallization occurred, the carbon steel showed stable and pure chemical compositions without any oxidation and interaction with the coating layer. In terms of the material aspects, these metallurgical inspection results suggest that the ZrO 2 coating performed well. (authors)

  5. Melting and evaporation analysis of the first wall in a water-cooled breeding blanket module under vertical displacement event by using the MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Geon-Woo [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Hyoung-Kyu, E-mail: chohk@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Im, Kihak [National Fusion Research Institute, 169-148 Gwahak-ro, Yuseong-gu, Daejeon 34133 (Korea, Republic of)

    2017-05-15

    Highlights: • Material phase change of first wall was simulated for vertical displacement event. • An in-house first wall module was developed to simulate melting and evaporation. • Effective heat capacity method and evaporation model were proposed. • MARS code was proposed to predict two-phase phenomena in coolant channel. • Phase change simulation was performed by coupling MARS and in-house module. - Abstract: Plasma facing components of tokamak reactors such as ITER or the Korean fusion demonstration reactor (K-DEMO) can be subjected to damage by plasma instabilities. Plasma disruptions like vertical displacement event (VDE) with high heat flux, can cause melting and vaporization of plasma facing materials and burnout of coolant channels. In this study, to simulate melting and vaporization of the first wall in a water-cooled breeding blanket under VDE, one-dimensional heat equations were solved numerically by using an in-house first wall module, including phase change models, effective heat capacity method, and evaporation model. For thermal-hydraulics, the in-house first wall analysis module was coupled with the nuclear reactor safety analysis code, MARS, to take advantage of its prediction capability for two-phase flow and critical heat flux (CHF) occurrence. The first wall was proposed for simulation according to the conceptual design of the K-DEMO, and the heat flux of plasma disruption with a value of 600 MW/m{sup 2} for 0.1 s was applied. The phase change simulation results were analyzed in terms of the melting and evaporation thicknesses and the occurrence of CHF. The thermal integrity of the blanket first wall is discussed to confirm whether the structural material melts for the given conditions.

  6. Ikaite crystals in melting sea ice - implications for pCO2 and pH levels in Arctic surface waters

    Science.gov (United States)

    Rysgaard, S.; Glud, R. N.; Lennert, K.; Cooper, M.; Halden, N.; Leakey, R. J. G.; Hawthorne, F. C.; Barber, D.

    2012-08-01

    A major issue of Arctic marine science is to understand whether the Arctic Ocean is, or will be, a source or sink for air-sea CO2 exchange. This has been complicated by the recent discoveries of ikaite (a polymorph of CaCO3·6H2O) in Arctic and Antarctic sea ice, which indicate that multiple chemical transformations occur in sea ice with a possible effect on CO2 and pH conditions in surface waters. Here, we report on biogeochemical conditions, microscopic examinations and x-ray diffraction analysis of single crystals from a melting 1.7 km2 (0.5-1 m thick) drifting ice floe in the Fram Strait during summer. Our findings show that ikaite crystals are present throughout the sea ice but with larger crystals appearing in the upper ice layers. Ikaite crystals placed at elevated temperatures disintegrated into smaller crystallites and dissolved. During our field campaign in late June, melt reduced the ice floe thickness by 0.2 m per week and resulted in an estimated 3.8 ppm decrease of pCO2 in the ocean surface mixed layer. This corresponds to an air-sea CO2 uptake of 10.6 mmol m-2 sea ice d-1 or to 3.3 ton km-2 ice floe week-1. This is markedly higher than the estimated primary production within the ice floe of 0.3-1.3 mmol m-2 sea ice d-1. Finally, the presence of ikaite in sea ice and the dissolution of the mineral during melting of the sea ice and mixing of the melt water into the surface oceanic mixed layer accounted for half of the estimated pCO2 uptake.

  7. Ikaite crystals in melting sea ice – implications for pCO2 and pH levels in Arctic surface waters

    Directory of Open Access Journals (Sweden)

    R. J. G. Leakey

    2012-08-01

    Full Text Available A major issue of Arctic marine science is to understand whether the Arctic Ocean is, or will be, a source or sink for air–sea CO2 exchange. This has been complicated by the recent discoveries of ikaite (a polymorph of CaCO3·6H2O in Arctic and Antarctic sea ice, which indicate that multiple chemical transformations occur in sea ice with a possible effect on CO2 and pH conditions in surface waters. Here, we report on biogeochemical conditions, microscopic examinations and x-ray diffraction analysis of single crystals from a melting 1.7 km2 (0.5–1 m thick drifting ice floe in the Fram Strait during summer. Our findings show that ikaite crystals are present throughout the sea ice but with larger crystals appearing in the upper ice layers. Ikaite crystals placed at elevated temperatures disintegrated into smaller crystallites and dissolved. During our field campaign in late June, melt reduced the ice floe thickness by 0.2 m per week and resulted in an estimated 3.8 ppm decrease of pCO2 in the ocean surface mixed layer. This corresponds to an air–sea CO2 uptake of 10.6 mmol m−2 sea ice d−1 or to 3.3 ton km−2 ice floe week−1. This is markedly higher than the estimated primary production within the ice floe of 0.3–1.3 mmol m−2 sea ice d−1. Finally, the presence of ikaite in sea ice and the dissolution of the mineral during melting of the sea ice and mixing of the melt water into the surface oceanic mixed layer accounted for half of the estimated pCO2 uptake.

  8. An approach to estimate the freshwater contribution from glacial melt and precipitation in East Greenland shelf waters using colored dissolved organic matter (CDOM)

    DEFF Research Database (Denmark)

    Stedmon, Colin; Granskog, Mats A.; Dodd, Paul A.

    2015-01-01

    Changes in the supply and storage of freshwater in the Arctic Ocean and its subsequent export to the North Atlantic can potentially influence ocean circulation and climate. In order to understand how the Arctic freshwater budget is changing and the potential impacts, it is important to develop......, and precipitation) and sea ice melt. We develop this approach further and investigate the use of an additional tracer, colored dissolved organic matter (CDOM), which is largely specific to freshwater originating from Arctic rivers. A robust relationship between the freshwater contribution from meteoric water...... processes (riverine input and sea ice formation), while previously, these waters where thought to be derived from open sea processes (cooling and sea ice formation) in the northern Barents and Kara Seas. In Greenlandic coastal waters the meteoric water contribution is influenced by Greenland ice sheet...

  9. 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

  10. Comparison of spray congealing and melt emulsification methods for the incorporation of the water-soluble salbutamol sulphate in lipid microparticles.

    Science.gov (United States)

    Scalia, Santo; Traini, Daniela; Young, Paul M; Di Sabatino, Marcello; Passerini, Nadia; Albertini, Beatrice

    2013-02-01

    Salbutamol sulphate is widely used as bronchodilator for the treatment of asthma. Its use is limited by the relatively short duration of action and hence sustained delivery of salbutamol sulphate offers potential benefits to patients. This study explores the preparation of lipid microparticles (LMs) as biocompatible carrier for the prolonged release of salbutamol sulphate. The LMs were produced using different lipidic materials and surfactants, by classical melt emulsification-based methods (oil-in-water and water-in-oil-in-water emulsions) and the spray congealing technique. For the LMs obtained by melt emulsification a lack of release modulation was observed. On the other hand, the sustained release of salbutamol sulphate was achieved with glyceryl behenate microparticles prepared by spray congealing. These LMs were characterized by scanning electron microscopy, X-ray diffractometry and differential scanning calorimetry. The drug loading was 4.72% (w/w). The particle size distribution measured by laser diffraction and electrical zone sensing was represented by a volume median diameter (Dv(50)) of 51.7-71.4 µm. Increasing the atomization air pressure from 4 to 8 bar produced a decrease of the Dv(50) to 12.7-17.5 µm. Incorporation of the hydrophilic salbutamol sulphate into LMs with sustained release characteristics was achieved by spray congealing.

  11. Volatile diffusion in silicate melts and its effects on melt inclusions

    Directory of Open Access Journals (Sweden)

    P. Scarlato

    2005-06-01

    Full Text Available A compendium of diffusion measurements and their Arrhenius equations for water, carbon dioxide, sulfur, fluorine, and chlorine in silicate melts similar in composition to natural igneous rocks is presented. Water diffusion in silicic melts is well studied and understood, however little data exists for melts of intermediate to basic compositions. The data demonstrate that both the water concentration and the anhydrous melt composition affect the diffusion coefficient of water. Carbon dioxide diffusion appears only weakly dependent, at most, on the volatilefree melt composition and no effect of carbon dioxide concentration has been observed, although few experiments have been performed. Based upon one study, the addition of water to rhyolitic melts increases carbon dioxide diffusion by orders of magnitude to values similar to that of 6 wt% water. Sulfur diffusion in intermediate to silicic melts depends upon the anhydrous melt composition and the water concentration. In water-bearing silicic melts sulfur diffuses 2 to 3 orders of magnitude slower than water. Chlorine diffusion is affected by both water concentration and anhydrous melt composition; its values are typically between those of water and sulfur. Information on fluorine diffusion is rare, but the volatile-free melt composition exerts a strong control on its diffusion. At the present time the diffusion of water, carbon dioxide, sulfur and chlorine can be estimated in silicic melts at magmatic temperatures. The diffusion of water and carbon dioxide in basic to intermediate melts is only known at a limited set of temperatures and compositions. The diffusion data for rhyolitic melts at 800°C together with a standard model for the enrichment of incompatible elements in front of growing crystals demonstrate that rapid crystal growth, greater than 10-10 ms-1, can significantly increase the volatile concentrations at the crystal-melt interface and that any of that melt trapped

  12. Influence of spatial discretization, underground water storage and glacier melt on a physically-based hydrological model of the Upper Durance River basin

    Science.gov (United States)

    Lafaysse, M.; Hingray, B.; Etchevers, P.; Martin, E.; Obled, C.

    2011-06-01

    SummaryThe SAFRAN-ISBA-MODCOU hydrological model ( Habets et al., 2008) presents severe limitations for alpine catchments. Here we propose possible model adaptations. For the catchment discretization, Relatively Homogeneous Hydrological Units (RHHUs) are used instead of the classical 8 km square grid. They are defined from the dilineation of hydrological subbasins, elevation bands, and aspect classes. Glacierized and non-glacierized areas are also treated separately. In addition, new modules are included in the model for the simulation of glacier melt, and retention of underground water. The improvement resulting from each model modification is analysed for the Upper Durance basin. RHHUs allow the model to better account for the high spatial variability of the hydrological processes (e.g. snow cover). The timing and the intensity of the spring snowmelt floods are significantly improved owing to the representation of water retention by aquifers. Despite the relatively small area covered by glaciers, accounting for glacier melt is necessary for simulating the late summer low flows. The modified model is robust over a long simulation period and it produces a good reproduction of the intra and interannual variability of discharge, which is a necessary condition for its application in a modified climate context.

  13. Surface Properties of Al-Functionalized Mesoporous MCM-41 and the Melting Behavior of Water in Al-MCM-41 Nanopores.

    Science.gov (United States)

    Sterczyńska, Angelina; Deryło-Marczewska, Anna; Zienkiewicz-Strzałka, Małgorzata; Śliwińska-Bartkowiak, Małgorzata; Domin, Kamila

    2017-10-24

    We report an experimental investigation of structural and adhesive properties for Al-containing mesoporous MCM-41 and MCM-41 surfaces. In this work, highly ordered hexagonal mesoporous structures of aluminosilica with two different Si/Al molar ratios equal to 50 and 80 and silica samples were studied; Al was incorporated into the MCM-41 structures using the direct synthesis method, with CTAB as a surfactant. The incorporation of aluminum was evidenced simultaneously without any change in the hexagonal arrangement of cylindrical mesopores. The porous materials were examined by techniques such as low-temperature nitrogen sorption, energy-dispersive spectroscopy, and scanning and transmission electron microscopy. Surface properties were determined through X-ray photoelectron spectroscopy, potentiometric titration, and static contact angle measurements. It was shown that an increase in surface acidity leads to an increase in the wetting energy of the surface. To investigate the influence of acidity on the confinement effects, the melting behavior of water in Al-MCM-41 and MCM-41 with the same pore size was determined by using dielectric relaxation spectroscopy and differential scanning calorimetry methods. We found that the melting-point depression of water in pores is larger in the functionalized pores than in pure silica pores of the same pore diameter.

  14. Pressure melting and ice skating

    Science.gov (United States)

    Colbeck, S. C.

    1995-10-01

    Pressure melting cannot be responsible for the low friction of ice. The pressure needed to reach the melting temperature is above the compressive failure stress and, if it did occur, high squeeze losses would result in very thin films. Pure liquid water cannot coexist with ice much below -20 °C at any pressure and friction does not increase suddenly in that range. If frictional heating and pressure melting contribute equally, the length of the wetted contact could not exceed 15 μm at a speed of 5 m/s, which seems much too short. If pressure melting is the dominant process, the water films are less than 0.08 μm thick because of the high pressures.

  15. Ikaite crystals in melting sea ice – implications for pCO2 and pH levels in Arctic surface waters

    DEFF Research Database (Denmark)

    Rysgaard, Søren; Glud, R.N.; Lennert, K.

    2012-01-01

    A major issue of Arctic marine science is to understand whether the Arctic Ocean is, or will be, a source or sink for air-sea CO 2 exchange. This has been complicated by the recent discoveries of ikaite (a polymorph of CaCO 3•6H 2O) in Arctic and Antarctic sea ice, which indicate that multiple...... chemical transformations occur in sea ice with a possible effect on CO 2 and pH conditions in surface waters. Here, we report on biogeochemical conditions, microscopic examinations and x-ray diffraction analysis of single crystals from a melting 1.7 km 2 (0.5-1 m thick) drifting ice floe in the Fram Strait...... during summer. Our findings show that ikaite crystals are present throughout the sea ice but with larger crystals appearing in the upper ice layers. Ikaite crystals placed at elevated temperatures disintegrated into smaller crystallites and dissolved. During our field campaign in late June, melt reduced...

  16. Pavement Snow Melting

    Energy Technology Data Exchange (ETDEWEB)

    Lund, John W.

    2005-01-01

    The design of pavement snow melting systems is presented based on criteria established by ASHRAE. The heating requirements depends on rate of snow fall, air temperature, relative humidity and wind velocity. Piping materials are either metal or plastic, however, due to corrosion problems, cross-linked polyethylene pipe is now generally used instead of iron. Geothermal energy is supplied to systems through the use of heat pipes, directly from circulating pipes, through a heat exchanger or by allowing water to flow directly over the pavement, by using solar thermal storage. Examples of systems in New Jersey, Wyoming, Virginia, Japan, Argentina, Switzerland and Oregon are presented. Key words: pavement snow melting, geothermal heating, heat pipes, solar storage, Wyoming, Virginia, Japan, Argentina, Klamath Falls.

  17. DEPENDENCY OF SULFATE SOLUBILITY ON MELT COMPOSITION AND MELT POLYMERIZATION

    International Nuclear Information System (INIS)

    JANTZEN, CAROL M.

    2004-01-01

    Sulfate and sulfate salts are not very soluble in borosilicate waste glass. When sulfate is present in excess it can form water soluble secondary phases and/or a molten salt layer (gall) on the melt pool surface which is purported to cause steam explosions in slurry fed melters. Therefore, sulfate can impact glass durability while formation of a molten salt layer on the melt pool can impact processing. Sulfate solubility has been shown to be compositionally dependent in various studies, (e.g. , B2O3, Li2O, CaO, MgO, Na2O, and Fe2O3 were shown to increase sulfate solubility while Al2O3 and SiO2 decreased sulfate solubility). This compositional dependency is shown to be related to the calculated melt viscosity at various temperatures and hence the melt polymerization

  18. Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

    International Nuclear Information System (INIS)

    Wang, Hesheng; Xu, Lifei; Chen, Weidong

    2016-01-01

    In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

  19. Design and implementation of visual inspection system handed in tokamak flexible in-vessel robot

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

    In-vessel viewing system (IVVS) is a fundamental tool among the remote handling systems for ITER, which is used to providing information on the status of the in-vessel components. The basic functional requirement of in-vessel visual inspection system is to perform a fast intervention with adequate optical resolution. In this paper, we present the software and hardware solution, which is designed and implemented for tokamak in-vessel viewing system that installed on end-effector of flexible in-vessel robot working under vacuum and high temperature. The characteristic of our in-vessel viewing system consists of two parts: binocular heterogeneous vision inspection tool and first wall scene emersion based augment virtuality. The former protected with water-cooled shield is designed to satisfy the basic functional requirement of visual inspection system, which has the capacity of large field of view and high-resolution for detection precision. The latter, achieved by overlaying first wall tiles images onto virtual first wall scene model in 3D virtual reality simulation system, is designed for convenient, intuitive and realistic-looking visual inspection instead of viewing the status of first wall only by real-time monitoring or off-line images sequences. We present the modular division of system, each of them in smaller detail, and go through some of the design choices according to requirements of in-vessel visual inspection task.

  20. Impact of polymer type on bioperformance and physical stability of hot melt extruded formulations of a poorly water soluble drug.

    Science.gov (United States)

    Mitra, Amitava; Li, Li; Marsac, Patrick; Marks, Brian; Liu, Zhen; Brown, Chad

    2016-05-30

    Amorphous solid dispersion formulations have been widely used to enhance bioavailability of poorly soluble drugs. In these formulations, polymer is included to physically stabilize the amorphous drug by dispersing it in the polymeric carrier and thus forming a solid solution. The polymer can also maintain supersaturation and promote speciation during dissolution, thus enabling better absorption as compared to crystalline drug substance. In this paper, we report the use of hot melt extrusion (HME) to develop amorphous formulations of a poorly soluble compound (FaSSIF solubility=1μg/mL). The poor solubility of the compound and high dose (300mg) necessitated the use of amorphous formulation to achieve adequate bioperformance. The effect of using three different polymers (HPMCAS-HF, HPMCAS-LF and copovidone), on the dissolution, physical stability, and bioperformance of the formulations was demonstrated. In this particular case, HPMCAS-HF containing HME provided the highest bioavailability and also had better physical stability as compared to extrudates using HPMCAS-LF and copovidone. The data demonstrated that the polymer type can have significant impact on the formulation bioperformance and physical stability. Thus a thorough understanding of the polymer choice is imperative when designing an amorphous solid dispersion formulation, such that the formulation provides robust bioperformance and has adequate shelf life. Copyright © 2016 Elsevier B.V. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  4. Electron beam melting of sponge titanium

    International Nuclear Information System (INIS)

    Kanayama, Hiroshi; Kusamichi, Tatsuhiko; Muraoka, Tetsuhiro; Onouye, Toshio; Nishimura, Takashi

    1991-01-01

    Fundamental investigations were done on electron beam (EB) melting of sponge titanium by using 80 kW EB melting furnace. Results obtained are as follows: (1) To increase the melting yield of titanium in EB melting of sponge titanium, it is important to recover splashed metal by installation of water-cooled copper wall around the hearth and to decrease evaporation loss of titanium by keeping the surface temperature of molten metal just above the melting temperature of titanium without local heating. (2) Specific power consumption of drip melting of pressed sponge titanium bar and hearth melting of sponge titanium are approximately 0.9 kWh/kg-Ti and 0.5-0.7 kWh/kg-Ti, respectively. (3) Ratios of the heat conducted to water-cooled mould in the drip melting and to water-cooled hearth in the hearth melting to the electron beam input power are 50-65% and 60-65%, respectively. (4) Surface defects of EB-melted ingots include rap which occurs when the EB output is excessively great, and transverse cracks when the EB output is excessively small. To prevent surface defects, the up-down withdrawal method is effective. (author)

  5. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K.I.; Kim, B.S.; Kim, D.H. [Korea Atomic Energy Research Inst., Thermal Hydraulic Safety Research, Taejon (Korea, Republic of)

    2001-07-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  6. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    International Nuclear Information System (INIS)

    Ahn, K.I.; Kim, B.S.; Kim, D.H.

    2001-01-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  8. Modified ITER In-Vessel Viewing System

    International Nuclear Information System (INIS)

    Ahola, H.; Heikkinen, V.; Keraenen, K.; Suomela, J.

    2001-01-01

    The original ITER In-Vessel Viewing System (IVVS) prototype (Proc. of the 20th SOFT, vol. 2 (1998) 1051), which demonstrates the feasibility of linear fibre arrays for ITER in-vessel viewing, has been modified. In order to reduce the viewing time and to improve the image quality the beam dispersing mirrors was replaced by a diffractive optics element (DOE), which enhanced the laser illumination considerably. The performance of the system was tested using various target surfaces: the results obtained clearly indicate its adequacy for in-vessel viewing. Mechanical damage on smooth metal surfaces (scratches etc.) can be easily distinguished and the viewing resolution at a distance of 2 m is better than 1 mm. The IVVS has been re-designed to be compatible with the new ITER-FEAT. A conceptual study which covers all the functions and subsystems required for viewing has been completed. These results will be used to further modify the prototype: items to be tested include horizontal probe operation and laser illumination with an optical fibre

  9. NET in-vessel vehicle system

    International Nuclear Information System (INIS)

    Jones, H.

    1991-02-01

    The CFFTP/Spar In-vessel Vehicle System concept for in-vessel remote maintenance of the NET/ITER machine is described. It comprises a curved deployable boom, a vehicle which can travel on the boom and an end effector or work unit mounted on the vehicle. The stowed boom, vehicle, and work unit are inserted via the equatorial access port of the torus. Following insertion the boom is deployed and locked in place. The vehicle may then travel along the boom to transport the work unit to any desired location. A novel feature of the concept is the deployable boom. When fully deployed, it closely resembles a conventional curved truss structure in configuration and characteristics. However, the joints of the truss structure are hinged so that it can fold into a compact package, of less than 20% of deployed volume for storage, transportation and insertion into the torus. A full-scale 2-metre long section of this boom was produced for demonstration purposes. As part of the concept definition the work unit for divertor handling was studied to demonstrate that large payloads could be manipulated within the confines of the torus using the in-vessel vehicle system. Principal advantages of the IVVS are its high load capacity and rigidity, low weight and stowed volume, simplicity of control and operation, and its relatively high speed of transportation

  10. GAREC analyses in Support of In-Vessel Retention Concept

    International Nuclear Information System (INIS)

    Azarian, G.; Gandrille, P.; Dumontet, A.; Grange; Barbier, F; Bellon, M.; Bordier, G.; Boulanger, F.; Cognet, G.; Gatt, J.M.; Humbert, J.M.; Laporte, T.; Lepareux, M.; Richard, P.; Robert, G.; Seiler, J.M.; Szabo, I.; Tourasse, M.; Valin, F.; Van Dorsselaere, J.P.

    1999-01-01

    The authors describe the analyses of the in-vessel retention capability which the GAREC group has performed for present and future French PWR designs. They present the reactor characteristics which are considered, describe the physical situations which are analysed and the relocation processes initiated by a corium flow, discuss the jet impacts, the debris formation and behaviour in the vessel lower head in a dry situation with absence of cooling, in wet situations in absence of external cooling, in wet situation with external cooling, in dry situation with external cooling. In this last case, they discuss the power dissipated in the corium, the molten salt behaviour, the heat flux distribution from the pool, the residual wall thickness, the heat flux distribution from the metal layer, the thermal-hydraulic aspects of water injection in the pool, the effects of crust instabilities, the external cooling, and the vessel mechanical behaviour. Then, they address the vapour explosion which may occur: mechanical loads leading to vessel failure in the cases of an eroded or non-eroded vessel, corium masses participating to the interaction (corium jets to the lower head, reflooding of corium pools with water). They finally briefly discuss the possible design improvements for in-vessel retention

  11. Melt spreading code assessment, modifications, and application to the EPR core catcher design

    International Nuclear Information System (INIS)

    Farmer, M.T.

    2009-01-01

    The Evolutionary Power Reactor (EPR) is under consideration by various utilities in the United States to provide base load electrical production, and as a result the design is undergoing a certification review by the U.S. Nuclear Regulatory Commission (NRC). The severe accident design philosophy for this reactor is based upon the fact that the projected power rating results in a narrow margin for in-vessel melt retention by external cooling of the reactor vessel. As a result, the design addresses ex-vessel core melt stabilization using a mitigation strategy that includes: (1) an external core melt retention system to temporarily hold core melt released from the vessel; (2) a layer of 'sacrificial' material that is admixed with the melt while in the core melt retention system; (3) a melt plug in the lower part of the retention system that, when failed, provides a pathway for the mixture to spread to a large core spreading chamber; and finally, (4) cooling and stabilization of the spread melt by controlled top and bottom flooding. The overall concept is illustrated in Figure 1.1. The melt spreading process relies heavily on inertial flow of a low-viscosity admixed melt to a segmented spreading chamber, and assumes that the melt mass will be distributed to a uniform height in the chamber. The spreading phenomenon thus needs to be modeled properly in order to adequately assess the EPR design. The MELTSPREAD code, developed at Argonne National Laboratory, can model segmented, and both uniform and nonuniform spreading. The NRC is thus utilizing MELTSPREAD to evaluate melt spreading in the EPR design. MELTSPREAD was originally developed to support resolution of the Mark I containment shell vulnerability issue. Following closure of this issue, development of MELTSPREAD ceased in the early 1990's, at which time the melt spreading database upon which the code had been validated was rather limited. In particular, the database that was utilized for initial validation consisted

  12. Protective coatings for in-vessel fusion devices

    International Nuclear Information System (INIS)

    Brossa, F.

    1984-01-01

    Coatings of Al/Si, SAP (Sintered Aluminium Powder), Al 2 O 3 , TiC (low-Z material) and Ta have been developed for in-vessel component protection. Anodic oxidation, vapor depositions, reactive sputtering, chemical vapor deposition (CVD) and plasma spray have been the coating formation methods studied. AISI 316, 310, 304, Inconel 600 and Mo were adopted as base materials. the coatings were characterized in terms of composition, structure and connection with the supporting material. The behavior of coatings under H + , D + and He + irradiation in the energy range 100 eV-8 keV was tested and compared to the solid massive samples. TiC and Ta coatings were tested with thermal shock under power density pulses of 1 kW/cm 2 generated by an electron beam gun. Temperature-dependence of the erosion of TiC by vacuum arcs in a magnetic field was also studied. TiC coatings have low sputtering values, good resistance to arcing and a high chemical stability. TiC and Ta, CVD and plasma spray coatings are thermal-shock resistant. High thermal loads produce cracks but no spalling. Destruction occurred only after melting of the base material. The plasma spray coating method seems to be most appropriate for developing remote handling applications in fusion devices. (orig.)

  13. ITER in-vessel system design and performance

    Science.gov (United States)

    Parker, R. R.

    2000-03-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

  14. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    1999-01-01

    This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

  15. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2001-01-01

    This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

  16. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2000-01-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events. (author)

  17. Sediment Lofting From Melt-Water Generated Turbidity Currents During Heinrich Events as a Tool to Assess Main Sediment Delivery Phases to Small Subpolar Ocean Basins

    Science.gov (United States)

    Hesse, R.

    2009-05-01

    equal density in the stratified water column, presumably the pycnocline, where they spread out laterally, even up-current, and generate interflows that deposit graded layers. The process is slow enough to allow incorporation into the graded layers of debris melting out of drifting icebergs. The hyperpycnal portion of individual discharge events that generated these currents might have had an estimated volume on the order of 103 km3x(1012 m3) which would have flowed for 10-15 days or less, assuming estimated discharge ranges for subglacial outburst floods of up to 106 m3/s. Turbidites are deposited much too fast to incorporate any substantial fraction of IRD, whereas deposition from lofted interflows may take months. The most likely candidates for the parent currents from which lofting occurred were the sandy flows that formed the sand abyssal plain The observed lofted depositional facies is exclusively found in Heinrich layers at distances of up to 300 km from the presumed terminus of the Hudson Strait ice stream. Through this stratigraphic relationship the lofted facies ties the main pulses of Late Pleistocene sediment supply in the Labrador Basin to Heinrich events. Heinrich events are known as Late Pleistocene ice-rafting episodes of unparalleled intensity in the North Atlantic that were associated with major melt-water discharge pulses and, as it appears now, also were the times of the main sediment delivery. Other potential basin candidates where lofting may have occurred are the Bering Sea and Maury Channel in North Atlantic.

  18. Hydrological scenarios for two selected Alpine catchments for the 21st century using a stochastic weather generator and enhanced process understanding for modelling of seasonal snow and glacier melt for improved water resources management

    Science.gov (United States)

    Strasser, Ulrich; Schneeberger, Klaus; Dabhi, Hetal; Dubrovsky, Martin; Hanzer, Florian; Marke, Thomas; Oberguggenberger, Michael; Rössler, Ole; Schmieder, Jan; Rotach, Mathias; Stötter, Johann; Weingartner, Rolf

    2016-04-01

    The overall objective of HydroGeM³ is to quantify and assess both water demand and water supply in two coupled human-environment mountain systems, i.e. Lütschine in Switzerland and Ötztaler Ache in Austria. Special emphasis is laid on the analysis of possible future seasonal water scarcity. The hydrological response of high Alpine catchments is characterised by a strong seasonal variability with low runoff in winter and high runoff in spring and summer. Climate change is expected to cause a seasonal shift of the runoff regime and thus it has significant impact on both amount and timing of the release of the available water resources, and thereof, possible future water conflicts. In order to identify and quantify the contribution of snow and ice melt as well as rain to runoff, streamflow composition will be analysed with natural tracers. The results of the field investigations will help to improve the snow and ice melt and runoff modules of two selected hydrological models (i.e. AMUNDSEN and WaSiM) which are used to investigate the seasonal water availability under current and future climate conditions. Together, they comprise improved descriptions of boundary layer and surface melt processes (AMUNDSEN), and of streamflow runoff generation (WaSiM). Future meteorological forcing for the modelling until the end of the century will be provided by both a stochastic multi-site weather generator, and downscaled climate model output. Both approches will use EUROCORDEX data as input. The water demand in the selected study areas is quantified for the relevant societal sectors, e.g. agriculture, hydropower generation and (winter) tourism. The comparison of water availability and water demand under current and future climate conditions will allow the identification of possible seasonal bottlenecks of future water supply and resulting conflicts. Thus these investigations can provide a quantitative basis for the development of strategies for sustainable water management in

  19. Water, lithium and trace element compositions of olivine from Lanzo South replacive mantle dunites (Western Alps): New constraints into melt migration processes at cold thermal regimes

    Science.gov (United States)

    Sanfilippo, Alessio; Tribuzio, Riccardo; Ottolini, Luisa; Hamada, Morihisa

    2017-10-01

    Replacive mantle dunites are considered to be shallow pathways for extraction of mantle melts from their source region. Dunites offer a unique possibility to unravel the compositional variability of the melts produced in the upper mantle, before mixing and crystal fractionation modify their original signature. This study includes a quantification of H2O, Li and trace elements (Ni, Mn, Co, Sc, V, Ti, Zr, Y and HREE) in olivine from large replacive dunite bodies (>20 m) within a mantle section exposed in the Western Italian Alps (Lanzo South ophiolite). On the basis of olivine, clinopyroxene and spinel compositions, these dunites were previously interpreted to be formed by melts with a MORB signature. Variations in Ni, Mn, Co and Ca contents in olivine from different dunite bodies suggested formation by different melt batches. The variable H2O and Li contents of these olivines agree with this idea. Compared to olivine from residual peridotites and olivine phenocrysts in MORB (both having H2O 1 ppm), the Lanzo South dunite olivine has high H2O (18-40 ppm) and low Li (0.35-0.83 ppm) contents. Geochemical modelling suggests that the dunite-forming melts were produced by low melting degrees of a mixed garnet-pyroxenite-peridotite mantle source, with a contribution of a garnet pyroxenite component variable from 20 to 80%. The Lanzo dunites experienced migration of melts geochemically enriched and mainly produced in the lowermost part of the melting region. Extraction of enriched melts through dunite channels are probably characteristic of cold thermal regimes, where low temperatures and a thick mantle lithosphere inhibit mixing with melts produced at shallower depths.

  20. Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

    International Nuclear Information System (INIS)

    Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

    2011-01-01

    In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)

  1. In-vessel core debris retention experiments. Final report

    International Nuclear Information System (INIS)

    1998-10-01

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

  2. Melting under shock compression

    International Nuclear Information System (INIS)

    Bennett, B.I.

    1980-10-01

    A simple model, using experimentally measured shock and particle velocities, is applied to the Lindemann melting formula to predict the density, temperature, and pressure at which a material will melt when shocked from room temperature and zero pressure initial conditions

  3. Thermodynamic study on the in-vessel corium - Application to the corium/concrete interaction

    International Nuclear Information System (INIS)

    Quaini, Andrea

    2015-01-01

    During a severe accident in a pressurised water reactor, the nuclear fuel can interact with the Zircaloy cladding, the neutronic absorber and the surrounding metallic structure forming a partially or completely molten mixture. The molten core can then interact with the reactor steel vessel forming a mixture called in-vessel corium. In the worst case, this mixture can pierce the vessel and pour onto the concrete underneath the reactor, leading the formation of the ex-vessel corium. Furthermore, depending on the considered scenario, the corium can be formed by a liquid phase or by two liquids, one metallic the other oxide. The objective of this thesis is the investigation of the thermodynamics of the prototypic in-vessel corium U-Pu-Zr- Fe-O. The approach used during the thesis is based on the CALPHAD method, which allows to obtain a thermodynamic model for this complex system starting from phase diagram and thermodynamic data. Heat treatments performed on the O-U-Zr system allowed to measure two tie-lines in the miscibility gap in the liquid phase at 2567 K. Furthermore, the liquidus temperatures of three Zr-enriched samples have been obtained by laser heating in collaboration with ITU. With the same laser heating technique, solidus temperatures have been obtained on the UO 2 -PuO 2 -ZrO 2 system. The influence of the reducing or oxidising on the melting behaviour of this system has been studied for the first time. The results show that the oxygen stoichiometry of these oxides strongly depends on the oxygen potential and on the metal composition of the samples. The miscibility gap in the liquid phase of the U-Zr-Fe-O system has been also observed. The whole set of experimental results with the literature data allowed to develop the thermodynamic model of the U-Pu-Zr-Fe-O system. Solidification path calculations have been performed for all the investigated samples to interpret the microstructures of the solidified samples. A good accordance has been obtained between

  4. Physico-Chemistry and Corium Properties for In-Vessel Retention

    International Nuclear Information System (INIS)

    Froment, K.; Seiler, J.M.; Gueneau, C.; Dauvois, V.; Barbier, F.; Bellon, M.; Tourasse, M.; Ducros, G.; Cognet, G.; Sudreau, F.

    1999-01-01

    This paper focuses on some important aspects of consequences of material behaviour and interactions on in-vessel retention capabilities. It discusses the behaviour of corium oxide mixtures at elevated temperatures (miscibility gap and density effects, separation due to density effects in the solid-liquid mixture according to the analysis of the Rasplav experiment results), and then the interaction between metallic layer and vessel wall (physical-chemical interaction of corium with the carbon steel vessel wall, migration of low melting point metallic elements in the solid vessel wall). It proposes a mode for the calculation of melt viscosity (liquid phase viscosity and viscosity in the solidification range), addresses the issue of barium release and residual power and of distribution of the residual power in an oxidic corium

  5. Dissolved iron in the Arctic shelf seas and surface waters of the central Arctic Ocean : Impact of Arctic river water and ice-melt

    NARCIS (Netherlands)

    Klunder, M. B.; Bauch, D.; Laan, P.; de Baar, H. J. W.; van Heuven, S.; Ober, S.

    2012-01-01

    Concentrations of dissolved (10 nM) in the bottom waters of the Laptev Sea shelf may be attributed to either sediment resuspension, sinking of brine or regeneration of DFe in the lower layers. A significant correlation (R-2 = 0.60) between salinity and DFe is observed. Using delta O-18, salinity,

  6. Experimental results for TiO2 melting and release using cold crucible melting

    International Nuclear Information System (INIS)

    Hong, S. W.; Min, B. T.; Park, I. G.; Kim, H. D.

    2000-01-01

    To simulate the severe accident phenomena using the real reactor material which melting point is about 2,800K, the melting and release method for materials with high melting point should be developed. This paper discusses the test results for TiO 2 materials using the cold crucible melting method to study the melting and release method of actual corium. To melt and release of few kg of TiO2, the experimental facility is manufactured through proper selection of design parameters such as frequency and capacity of R.F generator, crucible size and capacity of coolant. The melting and release of TiO 2 has been successfully performed in the cold crucible of 15cm in inner diameter and 30cm in height with 30kW RF power generator of 370 KHz. In the melt delivery experiment, about 2.6kg of molten TiO2, 60% of initial charged mass, is released. Rest of it is remained in the watercage in form of the rubble crust formed at the top of crucible and melt crust formed at the interface between the water-cage and melt. Especially, in the melt release test, the location of the working coil is important to make the thin crust at the bottom of the crucible

  7. Assessment of in-vessel corium retention in CPR1000

    International Nuclear Information System (INIS)

    Chen Xing; Zhang Shishun; Lin Jiming

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  10. Melting of Dense Sodium

    International Nuclear Information System (INIS)

    Gregoryanz, Eugene; Degtyareva, Olga; Hemley, Russell J.; Mao, Ho-kwang; Somayazulu, Maddury

    2005-01-01

    High-pressure high-temperature synchrotron diffraction measurements reveal a maximum on the melting curve of Na in the bcc phase at ∼31 GPa and 1000 K and a steep decrease in melting temperature in its fcc phase. The results extend the melting curve by an order of magnitude up to 130 GPa. Above 103 GPa, Na crystallizes in a sequence of phases with complex structures with unusually low melting temperatures, reaching 300 K at 118 GPa, and an increased melting temperature is observed with further increases in pressure

  11. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1991-01-01

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR [boiling water reactor] in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed

  12. A comparison of in-vessel behaviors between SFR and PWR under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sanggil; Cho, Cheon Hwey [ACT Co., Daejeon (Korea, Republic of); Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper aims to provide an easy guide for experts who know well the severe accident phenomenology of Pressurized Water Reactor (PWR) by comparing both reactor design concepts and in vessel behaviors under a postulated severe accident condition. This study only provides a preliminary qualitative comparison based on available literature. The PWR and SFR in-vessel design concepts and their effects under a postulate severe accident are investigated in this paper. Although this work is a preliminary study to compare the in-vessel behaviors for both PWR and SFR, it seems that there is no possibility to lead a significant core damage in the metal fuel SFR concept. In the oxide fuel SFR, there might be a chance to progress to the severe accident initiators such as the energetic reaction, flow blockage and so on.

  13. Lessons learnt from FARO/TERMOS corium melt quenching experiments

    Energy Technology Data Exchange (ETDEWEB)

    Magallon, D.; Huhtiniemi, I.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    The influence of melt quantity, melt composition, water depth and initial pressure on quenching is assessed on the basis of seven tests performed in various conditions in the TERMOS vessel of the FARO facility at JRC-Ispra. Tests involved UO{sub 2}-based melt quantities in the range 18-176 kg at a temperature of approximately 3000 K poured into saturated water. The results suggest that erosion of the melt jet column is an efficient contributor to the amount of break-up, and thus quenching, for large pours of corium melt. The presence of Zr metal in the melt induced a much more efficient quenching than in a similar test with no Zr metal, attributed to the oxidation of the Zr. Significant amounts of H{sub 2} were produced also in tests with pure oxidic melts (e.g. about 300 g for 157 kg melt). In the tests at 5.0 and 2.0 MPa good mixing with significant melt break-up and quenching was obtained during the penetration in the water. At 0.5 MPa, good penetration of the melt into the water could still be achieved, but a jump in the vessel pressurisation occurred when the melt contacted the bottom and part (5 kg) of the debris was re-ejected from the water. (author)

  14. Anticipating Central Asian Water Stress: Variation in River Flow Dependency on Melt Waters from Alpine to Plains in the Remote Tien Shan Range, Kyrgyzstan Using a Rapid Hydro Assessment Methodology

    Science.gov (United States)

    Hill, A. F.; Wilson, A. M.; Williams, M. W.

    2016-12-01

    The future of mountain water resources in High Asia is of high interest to water managers, development organizations and policy makers given large populations downstream reliant on snow and ice sourced river flow. Together with historical and cultural divides among ex-Soviet republics, a lack of central water management following the Soviet break-up has led to water stress as trans-boundary waters weave through and along borders. New upstream hydropower development, a thirsty downstream agricultural sector and a shrinking Aral Sea has led to increasing tension in the region. Despite these pressures and in contrast to eastern High Asia's Himalayan basins (Ganges, Brahmaputra), little attention has been given to western High Asia draining the Pamir and Tien Shan ranges (Syr Darya and Amu Darya basins) to better understand the hydrology of this vast and remote area. Difficult access and challenging terrain exacerbate challenges to working in this remote mountain region. As part of the Contributions to High Asia Runoff from Ice and Snow (CHARIS) project, we asked how does river flow source water composition change over an alpine-to-plains domain of Kyrgyzstan's Naryn River in the Syr Darya basin? In addition, what may the future hold for river flow in Central Asia given the differing responses of snow and ice to climate changes? Utilizing a Rapid Hydrologic Assessment methodology including a suite of pre-field mapping techniques we collected in situ water chemistry data at targeted, remote mountain sites over 450km of the Naryn River over an elevation gradient from glacial headwaters to the lower lying areas - places where people, hydropower and agriculture utilize water. Chemical and isotope tracers were used to separate stream flow to understand relative dependency on melt waters as the river moves downstream from glaciers and snow covered areas. This case study demonstrates a technique to acquire field data over large scales in remote regions that facilitates

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

    International Nuclear Information System (INIS)

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

    1999-04-01

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

  16. Experimental study on in-vessel debris coolability during severe accident

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  17. Model of interfacial melting

    DEFF Research Database (Denmark)

    Mouritsen, Ole G.; Zuckermann, Martin J.

    1987-01-01

    A two-dimensional model is proposed to describe systems with phase transitions which take place in terms of crystalline as well as internal degrees of freedom. Computer simulation of the model shows that the interplay between the two sets of degrees of freedom permits observation of grain-boundar......-boundary formation and interfacial melting, a nonequilibrium process by which the system melts at the boundaries of a polycrystalline domain structure. Lipid membranes are candidates for systems with pronounced interfacial melting behavior....

  18. Glacial melting in Himalaya

    Directory of Open Access Journals (Sweden)

    Kavita Tariyal

    2013-07-01

    Full Text Available Mountains are amongst the most flimsy environments on Earth. They are prosperous repositories of biodiversity, water and providers of ecosystem goods and services on which downstream communities, both regional and global, rely. The transport of atmospheric pollutants and climate-altering substances can significantly impact high mountain areas, which are generally considered “clean” regions. The snow glaciers of the Himalayas, considered the “third pole”, one of the largest stores of water on the planet and accelerated melting could have far-reaching effects, such as flooding in the short-term and water shortages in the long-term as the glaciers shrink. The data available on temperature in Himalayas indicate that warming during last 3-4 decades has been more than the global average over the last century. Some of the values indicate that the Himalayas are warming 5-6 times more than the global average. Mountain systems are seen globally as the prime sufferers from climate change. There is a severe gap in the knowledge of the short and long-term implications of the impact of climate change on water and hazards in the Himalayas, and their downstream river basins. Most studies have excluded the Himalayan region because of its extreme and complex topography and the lack of adequate rain gauge data. There is an urgent need to close the knowledge gap by establishing monitoring schemes for snow, ice and water; downscaling climate models; applying hydrological models to predict water availability; and developing basin wide scenarios, which also take water demand and socioeconomic development into account. Climate change induced hazards such as floods, landslides and droughts will impose considerable stresses on the livelihoods of mountain people and downstream populations. Enhancing resilience and promoting adaptation in mountain areas have thus become among the most important priorities of this decade. It is important to strengthen local

  19. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  20. Advanced in-vessel retention design for next generation risk management

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    In the TMI-2 accident, approximately twenty (20) tons of molten core material drained into the lower plenum. Early advanced light water reactor (LWR) designs assumed a lower head failure and incorporated various measures for ex-vessel accident mitigation. However,one of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 deg C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident/risk management strategies. As an advanced in-vessel design concept, this work presents the COrium Attack Syndrome Immunization Structures (COASIS) that are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in-vessel (COASISI) and ex-vessel (COASISO) are demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the Korean Standard Nuclear Power Plant (KSNPP) reactor. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. 15 refs., 5 figs., 1 tab. (Author)

  1. Distribution of the In-Vessel Diagnostics in ITER Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    González, Jorge, E-mail: Jorge.Gonzalez@iter.org [Rüecker Lypsa, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Clough, Matthew; Martin, Alex; Woods, Nick; Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France); Martinez, Gonzalo [Technical University Of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Stefan, Gicquel; Yunxing, Ma [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France)

    2017-01-15

    The ITER In-Vessel Diagnostics have been distributed around the In-Vessel shell to understand burning plasma physics and assist in machine operation. Each diagnostics component has its own requirements, constraints, and even exclusion among them for the highly complex In-Vessel environment. The size of the plasma, the requirement to be able to align the blanket system to the magnetic centre of the machine, the cooling requirements of the blanket system and the size of the pressure vessel itself all add to the difficulties of integrating these systems into the remaining space available. The available space for the cables inside the special trays (in-Vessel looms) is another constraint to allocate In-Vessel electrical sensors. Besides this, there are issues with the Assembly sequences and surface & volumetric neutron heating considerations that have imposed several additional restrictions.

  2. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  3. Core melt retention and cooling concept of the ERP

    Energy Technology Data Exchange (ETDEWEB)

    Weisshaeupl, H [SIEMENS/KWU, Erlangen (Germany); Yvon, M [Nuclear Power International, Paris (France)

    1996-12-01

    For the French/German European Pressurized Water Reactor (EPR) mitigative measures to cope with the event of a severe accident with core melt down are considered already at the design stage. Following the course of a postulated severe accident with reactor pressure vessel melt through one of the most important features of a future design must be to stabilize and cool the melt within the containment by dedicated measures. This measures should - as far as possible - be passive. One very promising solution for core melt retention seems to be a large enough spreading of the melt on a high temperature resistant protection layer with water cooling from above. This is the favorite concept for the EPR. In dealing with the retention of a molten core outside of the RPV several ``steps`` from leaving the RPV to finally stabilize the melt have to gone through. These steps are: collection of the melt; transfer of the melt; distribution of the melt; confining; cooling and stabilization. The technical features for the EPR solution of a large spreading of the melt are: Dedicated spreading chamber outside the reactor pit (area about 150 m{sup 2}); high temperature resistant protection layers (e.g. Zirconia bricks) at the bottom and part of the lateral structures (thus avoiding melt concrete interaction); reactor pit and spreading compartment are connected via a discharge channel which has a slope to the spreading area and is closed by a steel plate, which will resist the core melt for a certain time in order to allow a collection of the melt; the spreading compartments is connected with the In-Containment Refuelling Water Storage Tank (IRWST) with pipes for water flooding after spreading. These pipes are closed and will only be opened by the hot melt itself. It is shown how the course of the different steps mentioned above is processed and how each of these steps is automatically and passively achieved. (Abstract Truncated)

  4. Melt inclusions: Chapter 6

    Science.gov (United States)

    ,; Lowenstern, J. B.

    2014-01-01

    Melt inclusions are small droplets of silicate melt that are trapped in minerals during their growth in a magma. Once formed, they commonly retain much of their initial composition (with some exceptions) unless they are re-opened at some later stage. Melt inclusions thus offer several key advantages over whole rock samples: (i) they record pristine concentrations of volatiles and metals that are usually lost during magma solidification and degassing, (ii) they are snapshots in time whereas whole rocks are the time-integrated end products, thus allowing a more detailed, time-resolved view into magmatic processes (iii) they are largely unaffected by subsolidus alteration. Due to these characteristics, melt inclusions are an ideal tool to study the evolution of mineralized magma systems. This chapter first discusses general aspects of melt inclusions formation and methods for their investigation, before reviewing studies performed on mineralized magma systems.

  5. Ammonia-water mixtures at high pressures - Melting curves of ammonia dihydrate and ammonia monohydrate and a revised high-pressure phase diagram for the water-rich region. [in primordial solar system ices

    Science.gov (United States)

    Boone, S.; Nicol, M. F.

    1991-01-01

    The phase relations of some mixtures of ammonia and water are investigated to create a phase diagram in pressure-temperature-composition space relevant to the geophysical study of bodies in the outer solar system. The mixtures of NH3(x)H2O(1-x), where x is greater than 0.30 but less than 0.51, are examined at pressures and temperatures ranging from 0-6.5 GPa and 125-400 K, respectively. The ruby luminescence technique monitors the pressure and a diamond-anvil cell compresses the samples, and the phases are identified by means of normal- and polarized-light optical microscopy. The melting curve for NH3H2O(2) is described by the equation T = 176 + 60P - 8.5P squared for the ranges of 0.06-1.4 GPa and 179-243 K. The equation for NH3H2O is T = 194 + 37P - P squared, which represents a minor correction of a previous description by Johnson et al. (1985). Observed phase transitions are consistent with the high-pressure stability limit of NH3H2O(2), and the transition boundary is found to be linear.

  6. Melting Metal on a Playing Card

    Science.gov (United States)

    Greenslade, Thomas B., Jr.

    2016-01-01

    Many of us are familiar with the demonstration of boiling water in a paper cup held over a candle or a Bunsen burner; the ignition temperature of paper is above the temperature of 100°C at which water boils under standard conditions. A more dramatic demonstration is melting tin held in a playing card. This illustration is from Tissandier's book on…

  7. Processing and analysis techniques involving in-vessel material generation

    Science.gov (United States)

    Schabron, John F [Laramie, WY; Rovani, Jr., Joseph F.

    2012-09-25

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-01-01

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

  9. Active cooling system for Tokamak in-vessel operation manipulator

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Jianjun, E-mail: yuanjj@sjtu.edu.cn; Chen, Tan; Li, Fashe; Zhang, Weijun; Du, Liang

    2015-10-15

    Highlights: • We summarized most of the challenges of fusion devices to robot systems. • Propose an active cooling system to protect all of the necessary components. • Trial design test and theoretical analysis were conducted. • Overall implementation of the active cooling system was demonstrated. - Abstract: In-vessel operation/inspection is an indispensable task for Tokamak experimental reactor, for a robot/manipulator is more capable in doing this than human being with more precise motion and less risk of damaging the ambient equipment. Considering the demanding conditions of Tokamak, the manipulator should be adaptable to rapid response in the extreme conditions such as high temperature, vacuum and so on. In this paper, we propose an active cooling system embedded into such manipulator. Cameras, motors, gearboxes, sensors, and other mechanical/electrical components could then be designed under ordinary conditions. The cooling system cannot only be a thermal shield since the components are also heat sources in dynamics. We carry out a trial test to verify our proposal, and analyze the active cooling system theoretically, which gives a direction on the optimization by varying design parameters, components and distribution. And based on thermal sensors monitoring and water flow adjusting a closed-loop feedback control of temperature is added to the system. With the preliminary results, we believe that the proposal gives a way to robust and inexpensive design in extreme environment. Further work will concentrate on overall implementation and evaluation of this cooling system with the whole inspection manipulator.

  10. In-Vessel Retention via External Reactor Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Bachrata, Andrea [CTU in Prague, Faculty of nuclear sciences and physical engineering, V Holesovickach 2 180 00, Prague 8 (Czech republic)

    2008-07-01

    In-vessel (corium) retention (IVR) via external reactor pressure vessel (RPV) cooling is considered to be an effective severe accident management strategy for corium localisation and stabilisation. The main idea of IVR strategy consists in flooding the reactor cavity and transferring the decay heat through the wall of RPV to the recirculating water and than to the atmosphere of the containment of nuclear power plant. The aim of this strategy is to localise and to stabilise the corium inside the RPV. Not using this procedure could destroy the integrity of RPV and might cause the interaction of the corium with the concrete at the bed of the reactor cavity. Several experimental facilities and computer codes (MVITA, ASTEC module DIVA and CFD codes) were applied to simulate the IVR strategy for concrete reactor designs. The necessary technical modifications concerning the implementation of IVR concept were applied at the Loviisa NPP (VVER-440/V213). This strategy is also an important part of the advanced reactor designs AP600 and AP1000. (authors)

  11. Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Konovalenko, Alexander, E-mail: kono@kth.se; Karbojian, Aram, E-mail: karbojan@kth.se

    2017-04-01

    Highlights: • Steam explosion in stratified melt-coolant configuration is studied experimentally. • Different binary oxidic melt simulant materials were used. • Five spontaneous steam explosions were observed. • Instability of melt-coolant interface and formation of premixing layer was observed. • Explosion strength is influenced by melt superheat and water subcooling. - Abstract: Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.

  12. Development of ITER in-vessel viewing and metrology systems

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation ({approx}30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  13. Development of ITER in-vessel viewing and metrology systems

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation (∼30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  14. Force induced DNA melting

    International Nuclear Information System (INIS)

    Santosh, Mogurampelly; Maiti, Prabal K

    2009-01-01

    When pulled along the axis, double-strand DNA undergoes a large conformational change and elongates by roughly twice its initial contour length at a pulling force of about 70 pN. The transition to this highly overstretched form of DNA is very cooperative. Applying a force perpendicular to the DNA axis (unzipping), double-strand DNA can also be separated into two single-stranded DNA, this being a fundamental process in DNA replication. We study the DNA overstretching and unzipping transition using fully atomistic molecular dynamics (MD) simulations and argue that the conformational changes of double-strand DNA associated with either of the above mentioned processes can be viewed as force induced DNA melting. As the force at one end of the DNA is increased the DNA starts melting abruptly/smoothly above a critical force depending on the pulling direction. The critical force f m , at which DNA melts completely decreases as the temperature of the system is increased. The melting force in the case of unzipping is smaller compared to the melting force when the DNA is pulled along the helical axis. In the case of melting through unzipping, the double-strand separation has jumps which correspond to the different energy minima arising due to sequence of different base pairs. The fraction of Watson-Crick base pair hydrogen bond breaking as a function of force does not show smooth and continuous behavior and consists of plateaus followed by sharp jumps.

  15. In-vessel retention modeling capabilities in MAAP5

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  16. 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

  17. Endmembers of Ice Shelf Melt

    Science.gov (United States)

    Boghosian, A.; Child, S. F.; Kingslake, J.; Tedesco, M.; Bell, R. E.; Alexandrov, O.; McMichael, S.

    2017-12-01

    Studies of surface melt on ice shelves have defined a spectrum of meltwater behavior. On one end the storage of meltwater in persistent surface ponds can trigger ice shelf collapse as in the 2002 event leading to the disintegration of the Larsen B Ice Shelf. On the other, meltwater export by rivers can stabilize an ice shelf as was recently shown on the Nansen Ice Shelf. We explore this dichotomy by quantifying the partitioning between stored and transported water on two glaciers adjacent to floating ice shelves, Nimrod (Antarctica) and Peterman (Greenland). We analyze optical satellite imagery (LANDSAT, WorldView), airborne imagery (Operation IceBridge, Trimetrogon Aerial Phototography), satellite radar (Sentinel-1), and digital elevation models (DEMs) to categorize surface meltwater fate and map the evolution of ice shelf hydrology and topographic features through time. On the floating Peterman Glacier tongue a sizable river exports water to the ocean. The surface hydrology of Nimrod Glacier, geometrically similar to Peterman but with ten times shallower surface slope, is dominated by storage in surface lakes. In contrast, the Nansen has the same surface slope as Nimrod but transports water through surface rivers. Slope alone is not the sole control on ice shelf hydrology. It is essential to track the storage and transport volumes for each of these systems. To estimate water storage and transport we analyze high resolution (40 cm - 2 m) modern and historical DEMs. We produce historical (1957 onwards) DEMs with structure-from-motion photogrammetry. The DEMs are used to constrain water storage potential estimates of observed basins and water routing/transport potential. We quantify the total volume of water stored seasonally and interannually. We use the normalize difference water index to map meltwater extent, and estimate lake water depth from optical data. We also consider the role of stored water in subsurface aquifers in recharging surface water after

  18. Melting temperature of graphite

    International Nuclear Information System (INIS)

    Korobenko, V.N.; Savvatimskiy, A.I.

    2001-01-01

    Full Text: Pulse of electrical current is used for fast heating (∼ 1 μs) of metal and graphite specimens placed in dielectric solid media. Specimen consists of two strips (90 μm in thick) placed together with small gap so they form a black body model. Quasy-monocrystal graphite specimens were used for uniform heating of graphite. Temperature measurements were fulfilled with fast pyrometer and with composite 2-strip black body model up to melting temperature. There were fulfilled experiments with zirconium and tungsten of the same black body construction. Additional temperature measurements of liquid zirconium and liquid tungsten are made. Specific heat capacity (c P ) of liquid zirconium and of liquid tungsten has a common feature in c P diminishing just after melting. It reveals c P diminishing after melting in both cases over the narrow temperature range up to usual values known from steady state measurements. Over the next wide temperature range heat capacity for W (up to 5000 K) and Zr (up to 4100 K) show different dependencies of heat capacity on temperature in liquid state. The experiments confirmed a high quality of 2-strip black body model used for graphite temperature measurements. Melting temperature plateau of tungsten (3690 K) was used for pyrometer calibration area for graphite temperature measurement. As a result, a preliminary value of graphite melting temperature of 4800 K was obtained. (author)

  19. Melting of gold microclusters

    International Nuclear Information System (INIS)

    Garzon, I.L.; Jellinek, J.

    1991-01-01

    The transition from solid-like to liquid-like behavior in Au n , n=6, 7, 13, clusters is studied using molecular dynamics simulations. A Gupta-type potential with all-neighbour interactions is employed to incorporate n-body effects. The melting-like transition is described in terms of short-time averages of the kinetic energy per particle, root-mean-square bond length fluctuations and mean square displacements. A comparison between melting temperatures of Au n and Ni n clusters is presented. (orig.)

  20. Diffusion of hydrous species in model basaltic melt

    Science.gov (United States)

    Zhang, Li; Guo, Xuan; Wang, Qinxia; Ding, Jiale; Ni, Huaiwei

    2017-10-01

    Water diffusion in Fe-free model basaltic melt with up to 2 wt% H2O was investigated at 1658-1846 K and 1 GPa in piston-cylinder apparatus using both hydration and diffusion couple techniques. Diffusion profiles measured by FTIR are consistent with a model in which both molecular H2O (H2Om) and hydroxyl (OH) contribute to water diffusion. OH diffusivity is roughly 13% of H2Om diffusivity, showing little dependence on temperature or water concentration. Water diffusion is dominated by the motion of OH until total H2O (H2Ot) concentration reaches 1 wt%. The dependence of apparent H2Ot diffusivity on H2Ot concentration appears to be overestimated by a previous study on MORB melt, but H2Ot diffusivity at 1 wt% H2Ot in basaltic melt is still greater than those in rhyolitic to andesitic melts. The appreciable contribution of OH to water diffusion in basaltic melt can be explained by enhanced mobility of OH, probably associated with the development of free hydroxyl bonded with network-modifying cations, as well as higher OH concentration. Calculation based on the Nernst-Einstein equation demonstrates that OH may serve as an effective charge carrier in hydrous basaltic melt, which could partly account for the previously observed strong influence of water on electrical conductivity of basaltic melt.

  1. Fuel Rod Melt Progression Simulation Using Low-Temperature Melting Metal Alloy

    International Nuclear Information System (INIS)

    Seung Dong Lee; Suh, Kune Y.; GoonCherl Park; Un Chul Lee

    2002-01-01

    The TMI-2 accident and various severe fuel damage experiments have shown that core damage is likely to proceed through various states before the core slumps into the lower head. Numerous experiments were conducted to address when and how the core can lose its original geometry, what geometries are formed, and in what processes the core materials are transported to the lower plenum of the reactor pressure vessel. Core degradation progresses along the line of clad ballooning, clad oxidation, material interaction, metallic blockage, molten pool formation, melt progression, and relocation to the lower head. Relocation into the lower plenum may occur from the lateral periphery or from the bottom of the core depending upon the thermal and physical states of the pool. Determining the quantities and rate of molten material transfer to the lower head is important since significant amounts of molten material relocated to the lower head can threaten the vessel integrity by steam explosion and thermal and mechanical attack of the melt. In this paper the focus is placed on the melt flow regime on a cylindrical fuel rod utilizing the LAMDA (Lumped Analysis of Melting in Degrading Assemblies) facility at the Seoul National University. The downward relocation of the molten material is a combination of the external film flow and the internal pipe flow. The heater rods are 0.8 m long and are coated by a low-temperature melting metal alloy. The electrical internal heating method is employed during the test. External heating is adopted to simulate the exothermic Zircaloy-steam reaction. Tests are conducted in several quasi-steady-state conditions. Given the variable boundary conditions including the heat flux and the water level, observation is made for the melting location, progression, and the mass of molten material. Finally, the core melt progression model is developed from the visual inspection and quantitative analysis of the experimental data. As the core material relocates

  2. GLASS MELTING PHENOMENA, THEIR ORDERING AND MELTING SPACE UTILISATION

    Directory of Open Access Journals (Sweden)

    Němec L.

    2013-12-01

    Full Text Available Four aspects of effective glass melting have been defined – namely the fast kinetics of partial melting phenomena, a consideration of the melting phenomena ordering, high utilisation of the melting space, and effective utilisation of the supplied energy. The relations were defined for the specific melting performance and specific energy consumption of the glass melting process which involve the four mentioned aspects of the process and indicate the potentials of effective melting. The quantity “space utilisation” has been treated in more detail as an aspect not considered in practice till this time. The space utilisation was quantitatively defined and its values have been determined for the industrial melting facility by mathematical modelling. The definitions of the specific melting performance and specific energy consumption have been used for assessment of the potential impact of a controlled melt flow and high space utilisation on the melting process efficiency on the industrial scale. The results have shown that even the partial control of the melt flow, leading to the partial increase of the space utilisation, may considerably increase the melting performance, whereas a decrease of the specific energy consumption was determined to be between 10 - 15 %.

  3. Iter in vessel viewing system design and assessment activities

    Energy Technology Data Exchange (ETDEWEB)

    Neri, C., E-mail: carlo.neri@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Costa, P.; Ferri De Collibus, M.; Florean, M.; Mugnaini, G.; Pillon, M.; Pollastrone, F.; Rossi, P. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2011-10-15

    The In Vessel Viewing System (IVVS) is fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A prototype of a laser In Vessel Viewing and ranging System was developed and tested at ENEA laboratories in Frascati under EFDA task agreements, it is able to perform sub-millimetric bi-dimensional and three-dimensional images inside ITER during maintenance procedure allowing the evaluation of the state and damages of the in-vessel surface. The present prototype has been designed to operate under room conditions and starting from springtime 2009 a Grant with F4E is in progress for the design and the assessment of the IVVS system for ITER, keeping in account all the environmental conditions and constraints.

  4. Melt-processing method for radioactive solid wastes

    International Nuclear Information System (INIS)

    Kobayashi, Hiroaki

    1998-01-01

    Radioactive solid wastes are charged into a water-cooled type cold crucible induction melting furnace disposed in high frequency coils, and high frequency currents are supplied to high frequency coils which surround the melting furnace to melt the solid wastes by induction-heating. In this case, heat plasmas are jetted from above the solid wastes to the solid wastes to conduct initial heating to melt a portion of the solid wastes. Then, high frequency currents are supplied to the high frequency coils to conduct induction heating. According to this method, even when waste components of various kinds of materials are mixed, a portion of the solid wastes in the induction melting furnace can be melted by the initial heating by jetting heat plasmas irrespective of the kinds and the electroconductivity of the materials of the solid wastes. With such procedures, entire solid wastes in the furnace can be formed into a molten state uniformly and rapidly. (T.M.)

  5. Melt cooling by bottom flooding: The experiment CometPC-H3. Ex-vessel core melt stabilization research

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Cron, T.; Merkel, G.; Schmidt-Stiefel, S.; Tromm, W.; Wenz, T.

    2003-03-01

    The CometPC-H3 experiment was performed to investigate melt cooling by water addition to the bottom of the melt. The experiment was performed with a melt mass of 800 kg, 50% metal and 50% oxide, and 300 kW typical decay heat were simulated in the melt. As this was the first experiment after repair of the induction coil, attention was given to avoid overload of the induction coil and to keep the inductor voltage below critical values. Therefore, the height of the sacrificial concrete layer was reduced to 5 cm only, and the height of the porous concrete layers was also minimized to have a small distance and good coupling between heated melt and induction coil. After quite homogeneous erosion of the upper sacrificial concrete layer, passive bottom flooding started from the porous concrete after 220 s with 1.3 liter water/s. The melt was safely stopped, arrested and cooled. The porous, water filled concrete was only slightly attacked by the hot melt in the upper 25 mm of one sector of the coolant device. The peak cooling rate in the early contact phase of coolant water and melt was 4 MW/m 2 , and exceeded the decay heat by one order of magnitude. The cooling rate remarkably dropped, when the melt was covered by the penetrating water and a surface crust was formed. Volcanic eruptions from the melt during the solidification process were observed from 360 - 510 s and created a volcanic dome some 25 cm high, but had only minor effect on the generation of a porous structure, as the expelled melt solidified mostly with low porosity. Unfortunately, decay heat simulation in the melt was interrupted at 720 s by an incorrect safety signal, which excluded further investigation of the long term cooling processes. At that time, the melt was massively flooded by a layer of water, about 80 cm thick, and coolant water inflow was still 1 l/s. The melt had reached a stable situation: Downward erosion was stopped by the cooling process from the water filled, porous concrete layer. Top

  6. Multi-purpose deployer for ITER in-vessel maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

    2015-10-15

    Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

  7. MELT-IIIB: an updated version of the melt code

    International Nuclear Information System (INIS)

    Tabb, K.K.; Lewis, C.H.; O'Dell, L.D.; Padilla, A. Jr.; Smith, D.E.; Wilburn, N.P.

    1979-04-01

    The MELT series is a reactor modeling code designed to investigate a wide variety of hypothetical accident conditions, particularly the transient overpower sequence. MELT-IIIB is the latest in the series

  8. Study on superheat of TiAl melt during cold crucible levitation melting. TiAl no cold crucible levitation yokai ni okeru yoto kanetsudo no kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    Miwa, K.; Kobayashi, K.; Ninomiya, M. (Government Industrial Research Institute, Nagoya, Nagoya (Japan))

    1992-06-20

    Investigations were given on effects of test sample weights and sample positions in cold crucibles on superheat of melts when the intermetallic compound TiAl is melted using cold crucible levitation melting process, one of noncontaminated melting processes. The cold crucibles used in the experiment are a water-cooled copper crucible with an inner diameter of 42 mm and a length of 140 mm, into which a column-like ingot sample with an outer diameter of 32 mm (Al containing Ti at 33.5% by mass) was put and melted using the levitation melting. Comparisons and discussions were given on the relationship between sample weights and melt temperatures, the relationship between positions of the inserted samples and melt temperatures, and the state of contamination at melting of casts obtained from the melts resulted from the levitation melting and high-frequency melting poured into respective ceramic dies. Elevating the superheat temperature of the melts requires optimizing the sample weights and positions. Melt temperatures were measured using a radiation thermometer and a thermocouple, and the respective measured values were compared. 7 refs., 4 figs., 1 tab.

  9. Scrap uranium recycling via electron beam melting

    International Nuclear Information System (INIS)

    McKoon, R.

    1993-11-01

    A program is underway at the Lawrence Livermore National Laboratory (LLNL) to recycle scrap uranium metal. Currently, much of the material from forging and machining processes is considered radioactive waste and is disposed of by oxidation and encapsulation at significant cost. In the recycling process, uranium and uranium alloys in various forms will be processed by electron beam melting and continuously cast into ingots meeting applicable specifications for virgin material. Existing vacuum processing facilities at LLNL are in compliance with all current federal and state environmental, safety and health regulations for the electron beam melting and vaporization of uranium metal. One of these facilities has been retrofitted with an auxiliary electron beam gun system, water-cooled hearth, crucible and ingot puller to create an electron beam melt furnace. In this furnace, basic process R ampersand D on uranium recycling will be performed with the goal of eventual transfer of this technology to a production facility

  10. 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

  11. Plastic Melt Waste Compactor Flight Demonstrator Payload (PFDP), Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The PMWC Flight Demonstrator Payload is a trash dewatering and volume reduction system that uses heat melt compaction to remove nearly 100% of water from trash while...

  12. Experimental study of in-and-ex-vessel melt cooling during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sang Baik; Yoo, K J; Park, C K; Seok, S D; Park, R J; Yi, S J; Kang, K H; Ham, Y S; Cho, Y R; Kim, J H; Jeong, J H; Shin, K Y; Cho, J S; Kim, D H

    1997-07-01

    After code damage during a severe accident in a nuclear reactor, the degraded core has to be cooled down and the decay heat should be removed in order to cease the accident progression and maintain a stable state. The cooling of core melt is divided into in-vessel and ex-vessel cooling depending on the location of molten core which is dependent on the timing of vessel failure. Since the cooling mechanism varies with the conditions of molten core and surroundings and related phenomena, it contains many phenomenological uncertainties so far. In this study, an experimental study for verification of in-vessel corium cooling and several separate effect experiments for ex-vessel cooling are carried out to verify in- and ex-vessel cooling phenomena and finally to develop the accident management strategy and improve engineered reactor design for the severe accidents. SONATA-IV (Simulation of Naturally Arrested Thermal Attack in Vessel) program is set up for in-vessel cooling and a progression of the verification experiment has been done, and an integral verification experiment of the containment integrity for ex-vessel cooling is planned to be carried out based on the separate effect experiments performed in the first phase. First phase study of SONATA-IV is proof of principle experiment and it is composed of LALA (Lower-plenum Arrested Vessel Attack) experiment to find the gap between melt and the lower plenum during melt relocation and to certify melt quenching and CHFG (Critical Heat Flux in Gap) experiment to certify heat transfer mechanism in an artificial gap. As separate effect experiments for ex-vessel cooling, high pressure melt ejection experiment related to the initial condition for debris layer formation in the reactor cavity, crust formation and heat transfer experiment in the molten pool and molten core concrete interaction experiment are performed. (author). 150 refs., 24 tabs., 127 figs.

  13. In-vessel retention as a severe accident management strategy

    International Nuclear Information System (INIS)

    Theofanous, T.G.

    1999-01-01

    A global view of the in-vessel retention idea, as rendered recently in DOE/ID-10460 and DOE/ID-10541, is provided, with updates on on-going confirmatory work on fundamental aspects of the problem, and projections about potential future applications of the technology. The author's point of view on assessment of complex problems and scaling is emphasized. (author)

  14. Melting of polydisperse hard disks

    NARCIS (Netherlands)

    Pronk, S.; Frenkel, D.

    2004-01-01

    The melting of a polydisperse hard-disk system is investigated by Monte Carlo simulations in the semigrand canonical ensemble. This is done in the context of possible continuous melting by a dislocation-unbinding mechanism, as an extension of the two-dimensional hard-disk melting problem. We find

  15. Thermodynamics of Oligonucleotide Duplex Melting

    Science.gov (United States)

    Schreiber-Gosche, Sherrie; Edwards, Robert A.

    2009-01-01

    Melting temperatures of oligonucleotides are useful for a number of molecular biology applications, such as the polymerase chain reaction (PCR). Although melting temperatures are often calculated with simplistic empirical equations, application of thermodynamics provides more accurate melting temperatures and an opportunity for students to apply…

  16. Transient fuel melting

    International Nuclear Information System (INIS)

    Roche, L.; Schmitz, F.

    1982-10-01

    The observation of micrographic documents from fuel after a CABRI test leads to postulate a specific mode of transient fuel melting during a rapid nuclear power excursion. When reaching the melt threshold, the bands which are characteristic for the solid state are broken statistically over a macroscopic region. The time of maintaining the fuel at the critical enthalpy level between solid and liquid is too short to lead to a phase separation. A significant life-time (approximately 1 second) of this intermediate ''unsolide'' state would have consequences on the variation of physical properties linked to the phase transition solid/liquid: viscosity, specific volume and (for the irradiated fuel) fission gas release [fr

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

    International Nuclear Information System (INIS)

    Wider, H.

    2005-01-01

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

  18. Trajectory planning of tokamak flexible in-vessel inspection robot

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  19. Trajectory planning of tokamak flexible in-vessel inspection robot

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  20. Provision of reliable core cooling in vessel-type boiling reactors

    International Nuclear Information System (INIS)

    Alferov, N.S.; Balunov, B.F.; Davydov, S.A.

    1987-01-01

    Methods for providing reliable core cooling in vessel-type boiling reactors with natural circulation for heat supply are analysed. The solution of this problem is reduced to satisfaction of two conditions such as: water confinement over the reactor core necessary in case of an accident and confinement of sufficient coolant flow rate through the bottom cross section of fuel assemblies for some time. The reliable fuel element cooling under conditions of a maximum credible accident (brittle failure of a reactor vessel) is shown to be provided practically in any accident, using the safety vessel in combination with the application of means of standard operation and minimal composition and capacity of ECCS

  1. Some Elements of Equilibrium Diagrams for Systems of Iron with Water above 100 deg C and with Simple Chloride, Carbonate and Sulfate Melts

    International Nuclear Information System (INIS)

    Lewis, Derek

    1971-05-01

    Some aspects of molten salts relevant to the nuclear power industry are discussed briefly and an approach is presented to the theoretical description of the electrochemical thermodynamics of corrosion, mass-transport and deposition processes in power-plant working on the water-cycle, and in homogenous molten-salt reactors. Diagrams are introduced, based on the parameter pe instead of the usual redox potential, that are useful for illustrating equilibrium data for aqueous systems at elevated temperatures and for molten salts. Systems including iron in water and in the eutectic mixtures of lithium and sodium chloride, carbonate and sulfate, are taken as examples

  2. Some Elements of Equilibrium Diagrams for Systems of Iron with Water above 100 deg C and with Simple Chloride, Carbonate and Sulfate Melts

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Derek

    1971-05-15

    Some aspects of molten salts relevant to the nuclear power industry are discussed briefly and an approach is presented to the theoretical description of the electrochemical thermodynamics of corrosion, mass-transport and deposition processes in power-plant working on the water-cycle, and in homogenous molten-salt reactors. Diagrams are introduced, based on the parameter pe instead of the usual redox potential, that are useful for illustrating equilibrium data for aqueous systems at elevated temperatures and for molten salts. Systems including iron in water and in the eutectic mixtures of lithium and sodium chloride, carbonate and sulfate, are taken as examples

  3. Numerical simulation of fragmentation of hot metal and oxide melts with the computer code IVA3

    International Nuclear Information System (INIS)

    Mussa, S.; Tromm, W.

    1994-01-01

    The phenomena of fragmentation of melts caused by water-inlet from the bottom with the computer code IVA3/11,12,13/ are investigated. With the computer code IVA3 three-component-multiphase flows can be numerically simulated. Two geometrical models are used. Both consist of a cylindrical vessel for water lying beneath a cylindrical vessel for melt. The vessels are connected to each other through a hole. Steel and UO 2 melts are. The following parameters were varied: the type of the melt (steel,UO 2 ), the water supply pressure and the geometry of the hole in the bottom plate through which the water and melt vessels are connected. As results of the numerical simulations temperature and pressure versus time curves are plotted. Additionally the volume flow rates and the volume fractions of the various phases in the vessels and the increase in surface and enthalpy of the melt during the time of simulation are depicted. With steel melts the rate of fragmentation increases with increasing water pressure and melt temperature, whereby stable channels are formed in the melt layer showing a very low flow resistance for steam. With UO 2 the formations of channels are also observed. However, these channels are not so stable that they eventually break apart and lead to the fragmentation of the UO 2 melt in drops. The fragmentation of the steel melt in water vessel is less than that of UO 2 . No essential solidification of the melt is observed in the respective duration of the simulations. However, a small drop in the melt temperature is observed. With a slight or no water pressure the melt flows from the upper vessel into the water vessel via the connecting hole. The processes take place in a very slow manner and with such a low steam production so that despite the occuring pressure peaks no sign of steam explosions could be observed. (orig./HP) [de

  4. Emerging melt quality control solution technologies for aluminium melt

    Directory of Open Access Journals (Sweden)

    Arturo Pascual, Jr

    2009-11-01

    Full Text Available The newly developed “MTS 1500” Melt Treatment System is performing the specifi cally required melt treatment operations like degassing, cleaning, modification and/or grain refinement by an automated process in one step and at the same location. This linked process is saving time, energy and metal losses allowing - by automated dosage of the melt treatment agents - the production of a consistent melt quality batch after batch. By linking the MTS Metal Treatment System with sensors operating on-line in the melt, i.e., with a hydrogen sensor “Alspek H”, a fully automated control of parts of the process chain like degassing is possible. This technology does guarantee a pre-specifi ed and documented melt quality in each melt treatment batch. Furthermore, to ensure that castings are consistent and predictable there is a growing realization that critical parameters such as metal cleanliness must be measured prior to casting. There exists accepted methods for measuring the cleanliness of an aluminum melt but these can be both slow and costly. A simple, rapid and meaningful method of measuring and bench marking the cleanliness of an aluminum melt has been developed to offer the foundry a practical method of measuring melt cleanliness. This paper shows the structure and performance of the integrated MTS melt treatment process and documents achieved melt quality standards after degassing, cleaning, modifi cation and grain refi nement operations under real foundry conditions. It also provides an insight on a melt cleanliness measuring device “Alspek MQ” to provide foundry men better tools in meeting the increasing quality and tighter specifi cation demand from the industry.

  5. Polycarboxylic acids as network modifiers for water durability improvement of inorganic-organic hybrid tin-silico-phosphate low-melting glasses

    International Nuclear Information System (INIS)

    Menaa, Bouzid; Mizuno, Megumi; Takahashi, Masahide; Tokuda, Yomei; Yoko, Toshinobu

    2006-01-01

    We investigated the water durability of the inorganic-organic hybrid tin-silico-phosphate glasses Me 2 SiO-SnO-P 2 O 5 (Me designs the organic methyl group) doped with organic acids (salicylic acid (SA), tartaric acid (TA), citric acid (Canada) and butane tetracarboxylic acid (BTCA)) containing one or more of carboxylic groups per molecule. The structure, thermal properties and durability of the final glasses obtained via a non-aqueous acid-base reaction were discussed owing to the nature and the concentration of the acid added. 29 Si magic angle spinning (MAS) NMR and 31 P MAS NMR spectra, respectively, showed clearly a modification of the network in the host glass matrix of the Me 2 SiO-SnO-P 2 O 5 system. The polycondensation enhancement to form -P-O-Si-O-P- linkages (PSP) and the increase of the Q 2 unit (two bridging oxygens per phosphorus atom) over the Q 3 unit (three bridging oxygens per phosphorus atom) as a function of the acid in the order SA 2 SiO-SnO-P 2 O 5 matrix. In addition, this structural change is accompanied by a decrease of the coefficient of thermal expansion and an increase of the water durability of the glasses with the acids containing a large number of carboxylic groups per molecule. The presence of carboxylic groups of the acid acting as network modifier may retard the movement of water molecules through the glasses due to the steric hindrance strengthening the PSP connections in a chain-like structure

  6. In-vessel maintenance concepts for tokamak fusion reactors

    International Nuclear Information System (INIS)

    Kelly, V.P.; Berger, J.D.; Yount, J.A.

    1983-01-01

    Concepts for rail-mounted and guided in-vessel handling machines (IVM) for remote maintenance inside tokamak fusion reactors are described. The IVM designs are based on concepts for tethered remotely operated vehicles and feature the use of multiple manipulator arms for remote handling and remote-controlled TV cameras for remote viewing. The concepts include IVMs for both single or dual rail systems located in the top or bottom of the reactor vessel

  7. Proposal of In-vessel corium retention concept for Paks NPP

    International Nuclear Information System (INIS)

    Elter, J.; Toth, E.; Matejovic, P.

    2011-01-01

    The in-vessel corium retention (IVR) via external reactor vessel cooling (ERVC) seems to be a promising severe accident management strategy not only for new generation of advanced PWRs, but also for VVER-440/V213 reactors, which were designed several years ago. The basic idea of in-vessel retention of corium is to prevent RPV failure by flooding the reactor cavity so that the reactor pressure vessel is submerged in water up to its support structures, and thus the decay heat can be transferred from the corium pool through the vessel wall and into the water surrounding the vessel. An IVR concept with simple ECVR loop based only on minor modifications of existing plant technology was proposed for the Paks Nuclear Power Plant. 2 severe accident (LB and SB LOCA) without availability of HP and LP safety injection in power upgrade (108%) conditions were simulated using the ASTEC code. The analyses show that the proposed solution is effective in preserving RPV integrity in the case of severe accident. Possible uncertainties in code predictions are covered by the applied conservative assumptions

  8. Method of melting solid waste

    International Nuclear Information System (INIS)

    Ootsuka, Katsuyuki; Mizuno, Ryokichi; Kuwana, Katsumi; Sawada, Yoshihisa; Komatsu, Fumiaki.

    1982-01-01

    Purpose: To enable the volume reduction treatment of a HEPA filter containing various solid wastes, particularly acid digestion residue, or an asbestos separator at a relatively low temperature range. Method: Solid waste to be heated and molten is high melting point material treated by ''acid digestion treatment'' for treating solid waste, e.g. a HEPA filter or polyvinyl chloride, etc. of an atomic power facility treated with nitric acid or the like. When this material is heated and molten by an electric furnace, microwave melting furnace, etc., boron oxide, sodium boride, sodium carbonate, etc. is added as a melting point lowering agent. When it is molten in this state, its melting point is lowered, and it becomes remarkably fluid, and the melting treatment is facilitated. Solidified material thus obtained through the melting step has excellent denseness and further large volume reduction rate of the solidified material. (Yoshihara, H.)

  9. Petrological Geodynamics of Mantle Melting II. AlphaMELTS + Multiphase Flow: Dynamic Fractional Melting

    Science.gov (United States)

    Tirone, Massimiliano

    2018-03-01

    In this second installment of a series that aims to investigate the dynamic interaction between the composition and abundance of the solid mantle and its melt products, the classic interpretation of fractional melting is extended to account for the dynamic nature of the process. A multiphase numerical flow model is coupled with the program AlphaMELTS, which provides at the moment possibly the most accurate petrological description of melting based on thermodynamic principles. The conceptual idea of this study is based on a description of the melting process taking place along a 1-D vertical ideal column where chemical equilibrium is assumed to apply in two local sub-systems separately on some spatial and temporal scale. The solid mantle belongs to a local sub-system (ss1) that does not interact chemically with the melt reservoir which forms a second sub-system (ss2). The local melt products are transferred in the melt sub-system ss2 where the melt phase eventually can also crystallize into a different solid assemblage and will evolve dynamically. The main difference with the usual interpretation of fractional melting is that melt is not arbitrarily and instantaneously extracted from the mantle, but instead remains a dynamic component of the model, hence the process is named dynamic fractional melting (DFM). Some of the conditions that may affect the DFM model are investigated in this study, in particular the effect of temperature, mantle velocity at the boundary of the mantle column. A comparison is made with the dynamic equilibrium melting (DEM) model discussed in the first installment. The implications of assuming passive flow or active flow are also considered to some extent. Complete data files of most of the DFM simulations, four animations and two new DEM simulations (passive/active flow) are available following the instructions in the supplementary material.

  10. Logistics Reduction: Heat Melt Compactor

    Data.gov (United States)

    National Aeronautics and Space Administration — The Advanced Exploration Systems (AES) Logistics Reduction (LR) project Heat Melt Compactor (HMC) technology is a waste management technology. Currently, there are...

  11. Melting in trivalent metal chlorides

    International Nuclear Information System (INIS)

    Saboungi, M.L.; Price, D.L.; Scamehorn, C.; Tosi, M.P.

    1990-11-01

    We report a neutron diffraction study of the liquid structure of YCl 3 and combine the structural data with macroscopic melting and transport data to contrast the behaviour of this molten salt with those of SrCl 2 , ZnCl 2 and AlCl 3 as prototypes of different melting mechanisms for ionic materials. A novel melting mechanism for trivalent metal chlorides, leading to a loose disordered network of edge-sharing octahedral units in the liquid phase, is thereby established. The various melting behaviours are related to bonding character with the help of Pettifor's phenomenological chemical scale. (author). 25 refs, 4 figs, 3 tabs

  12. Melting of contaminated metallic waste

    International Nuclear Information System (INIS)

    Lee, Y.-S.; Cheng, S.-Y.; Kung, H.-T.; Lin, L.-F.

    2004-01-01

    Approximately 100 tons of contaminated metallic wastes were produced each year due to maintenance for each TPC's nuclear power reactor and it was roughly estimated that there will be 10,000 tons of metallic scraps resulted from decommissioning of each reactor in the future. One means of handling the contaminated metal is to melt it. Melting process owns not only volume reduction which saves the high cost of final disposal but also resource conservation and recycling benefits. Melting contaminated copper and aluminum scraps in the laboratory scale have been conducted at INER. A total of 546 kg copper condenser tubes with a specific activity of about 2.7 Bq/g was melted in a vacuum induction melting facility. Three types of products, ingot, slag and dust were derived from the melting process, with average activities of 0.10 Bq/g, 2.33 Bq/g and 84.3 Bq/g respectively. After the laboratory melting stage, a pilot plant with a 500 kg induction furnace is being designed to melt the increasingly produced contaminated metallic scraps from nuclear facilities and to investigate the behavior of different radionuclides during melting. (author)

  13. Comparison of a novel spray congealing procedure with emulsion-based methods for the micro-encapsulation of water-soluble drugs in low melting point triglycerides.

    Science.gov (United States)

    McCarron, Paul A; Donnelly, Ryan F; Al-Kassas, Rasil

    2008-09-01

    The particle size characteristics and encapsulation efficiency of microparticles prepared using triglyceride materials and loaded with two model water-soluble drugs were evaluated. Two emulsification procedures based on o/w and w/o/w methodologies were compared to a novel spray congealing procedure. After extensive modification of both emulsification methods, encapsulation efficiencies of 13.04% tetracycline HCl and 11.27% lidocaine HCl were achievable in a Witepsol-based microparticle. This compares to much improved encapsulation efficiencies close to 100% for the spray congealing method, which was shown to produce spherical particles of approximately 58 microm. Drug release studies from a Witepsol formulation loaded with lidocaine HCl showed a temperature-dependent release mechanism, which displayed diffusion-controlled kinetics at temperatures approximately 25 degrees C, but exhibited almost immediate release when triggered using temperatures close to that of skin. Therefore, such a system may find application in topical semi-solid formulations, where a temperature-induced burst release is preferred.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  15. Melting method for miscellaneous radioactive solid waste and melting furnace

    International Nuclear Information System (INIS)

    Osaki, Toru; Furukawa, Hirofumi; Uda, Nobuyoshi; Katsurai, Kiyomichi

    1998-01-01

    A vessel containing miscellaneous solid wastes is inserted in a crucible having a releasable material on the inner surface, they are induction-heated from the outside of the crucible by way of low temperature heating coils to melt low melting point materials in the miscellaneous wastes within a temperature range at which the vessel does not melt. Then, they are induction-heated by way of high temperature heating coils to melt the vessel and not yet melted materials, those molten materials are cooled, solidified molten material and the releasable material are taken out, and then the crucible is used again. Then, the crucible can be used again, so that it can be applied to a large scaled melting furnace which treats wastes by a unit of drum. In addition, since the cleaning of the used crucible and the application of the releasable material can be conducted without interrupting the operation of the melting furnace, the operation cycle of the melting furnace can be shortened. (N.H.)

  16. Cleaning, disassembly, and requalification of the FFTF in vessel handling machine

    International Nuclear Information System (INIS)

    Coops, W.J.

    1977-10-01

    The Engineering Model In Vessel Handling Machine (IVHM) was successfully removed, cleaned, disassembled, inspected, reassembled and reinstalled into the sodium test vessel at Richland, Washington. This was the first time in the United States a full size operational sodium wetted machine has been cleaned by the water vapor nitrogen process and requalified for operation. The work utilized an atmospheric control system during removal, a tank type water vapor nitrogen cleaning system and an open ''hands on'' disassembly and assembly stand. Results of the work indicate the tools, process and equipment are adequate for the non-radioactive maintenance sequence. Additionally, the work proves that a machine of this complexity can be successfully cleaned, maintained and re-used without the need to replace a large percentage of the sodium wetted parts

  17. Waste glass melting stages

    International Nuclear Information System (INIS)

    Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

    1994-01-01

    Three simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C to 1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentru Karlsruhe (KfK) in Germany were used. The samples were thin sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. The behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied. 2 refs., 8 tabs

  18. Waste glass melting stages

    International Nuclear Information System (INIS)

    Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

    1993-04-01

    Three different simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentrum Karlsruhe (KfK) in Germany were used. The samples were thin-sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. Behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied

  19. Estimation of Melt Ponds over Arctic Sea Ice using MODIS Surface Reflectance Data

    Science.gov (United States)

    Ding, Y.; Cheng, X.; Liu, J.

    2017-12-01

    Melt ponds over Arctic sea ice is one of the main factors affecting variability of surface albedo, increasing absorption of solar radiation and further melting of snow and ice. In recent years, a large number of melt ponds have been observed during the melt season in Arctic. Moreover, some studies have suggested that late spring to mid summer melt ponds information promises to improve the prediction skill of seasonal Arctic sea ice minimum. In the study, we extract the melt pond fraction over Arctic sea ice since 2000 using three bands MODIS weekly surface reflectance data by considering the difference of spectral reflectance in ponds, ice and open water. The preliminary comparison shows our derived Arctic-wide melt ponds are in good agreement with that derived by the University of Hamburg, especially at the pond distribution. We analyze seasonal evolution, interannual variability and trend of the melt ponds, as well as the changes of onset and re-freezing. The melt pond fraction shows an asymmetrical growth and decay pattern. The observed melt ponds fraction is almost within 25% in early May and increases rapidly in June and July with a high fraction of more than 40% in the east of Greenland and Beaufort Sea. A significant increasing trend in the melt pond fraction is observed for the period of 2000-2017. The relationship between melt pond fraction and sea ice extent will be also discussed. Key Words: melt ponds, sea ice, Arctic

  20. Experimental studies on melt spreading, bubbling heat transfer, and coolant layer boiling

    International Nuclear Information System (INIS)

    Greene, G.A.; Finfrock, C.; Klages, J.; Schwarz, C.E.; Burson, S.B.

    1988-01-01

    Melt spreading studies have been undertaken to investigate the extent to which molten core debris may be expected to spread under gravity forces in a BWR drywell geometry. The objectives are to determine the extent of melt spreading as a function of melt mass,melt superheat, and water depth. These studies will enable an objective determination of whether or not core debris can spread up to and contact containment structures or boundaries upon vessel failure. Results indicate that the most important variables are the melt superheat and the water depth. Studies have revealed five distinct regimes of melt spreading ranging from hydrodynamically-limited to heat transfer-limited. A single parameter dimensionless correlation is presented which identified the spreading regime and allows for mechanistic calculation of the average thickness to which the melt will spread. 7 refs., 12 figs

  1. Rhenium corrosion in chloride melts

    International Nuclear Information System (INIS)

    Stepanov, A.D.; Shkol'nikov, S.N.; Vetyukov, M.M.

    1989-01-01

    The results investigating rhenium corrosion in chloride melts containing sodium, potassium and chromium ions by a gravimetry potentials in argon atmosphere in a sealing quarth cell are described. Rhenium corrosion is shown to be rather considerable in melts containing CrCl 2 . The value of corrosion rate depending on temperature is determined

  2. UNCONSTRAINED MELTING AND SOLIDIFICATION INSIDE ...

    African Journals Online (AJOL)

    2015-09-01

    Sep 1, 2015 ... There is a large number of experimental and numerical works on melting and solidification of PCM[6-10], and also its usage as thermal management in building [11-14], electronic devices [15-16] and solar energy. [17-20].Most investigated geometries in melting and freezing process are sphere (spherical.

  3. The JET high temperature in-vessel inspection system

    International Nuclear Information System (INIS)

    Businaro, T.; Cusack, R.; Calbiati, L.; Raimondi, T.

    1989-01-01

    The JET In-vessel Inspection System (IVIS) has been enhanced for operation under the following nominal conditions: vacuum vessel at 350 degC; vacuum vessel evacuated (∼10 -9 mbar); radiation dose during D-T phase 10 rads. The target resolution of the pictures is 2 mm at 5 m distance and tests on radiation resistance of the IVIS system are being carried out. Since June 1988, the new system is installed in the JET machine and the first inspections of the intire vessel at 250 degC have been satisfactory done. (author). 3 refs.; 6 figs.; 1 tab

  4. Evaluation of microwave cavity gas sensor for in-vessel monitoring of dry cask storage systems

    Science.gov (United States)

    Bakhtiari, S.; Gonnot, T.; Elmer, T.; Chien, H.-T.; Engel, D.; Koehl, E.; Heifetz, A.

    2018-04-01

    Results are reported of research activities conducted at Argonne to assess the viability of microwave resonant cavities for extended in-vessel monitoring of dry cask storage system (DCSS) environment. One of the gases of concern to long-term storage in canisters is water vapor, which appears due to evaporation of residual moisture from incompletely dried fuel assembly. Excess moisture could contribute to corrosion and deterioration of components inside the canister, which would in turn compromise maintenance and safe transportation of such systems. Selection of the sensor type in this work was based on a number of factors, including good sensitivity, fast response time, small form factor and ruggedness of the probing element. A critical design constraint was the capability to mount and operate the sensor using the existing canister penetrations-use of existing ports for thermocouple lances. Microwave resonant cavities operating at select resonant frequency matched to the rotational absorption line of the molecule of interest offer the possibility of highly sensitive detection. In this study, two prototype K-band microwave cylindrical cavities operating at TE01n resonant modes around the 22 GHz water absorption line were developed and tested. The sensors employ a single port for excitation and detection and a novel dual-loop inductive coupling for optimized excitation of the resonant modes. Measurement of the loaded and unloaded cavity quality factor was obtained from the S11 parameter. The acquisition and real-time analysis of data was implemented using software based tools developed for this purpose. The results indicate that the microwave humidity sensors developed in this work could be adapted to in-vessel monitoring applications that require few parts-per-million level of sensitivity. The microwave sensing method for detection of water vapor can potentially be extended to detection of radioactive fission gases leaking into the interior of the canister through

  5. Evaluation Plan on In-vessel Source Term in PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Won; Ha, Kwi-Seok; Ahn, Sang June; Lee, Kwi Lim; Jeong, Taekyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. The in-vessel STs of PGSFR will be estimated using the methodology of ANL-ART-38 report in additional to 4S methodology. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the realistic burnup conditions. ST in the release from the core to primary sodium is calculated by using the assumption of ANL methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using equation and experimental materials.

  6. Design features of the KSTAR in-vessel control coils

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-06-15

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

  7. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  8. Evaluation Plan on In-vessel Source Term in PGSFR

    International Nuclear Information System (INIS)

    Lee, Seung Won; Ha, Kwi-Seok; Ahn, Sang June; Lee, Kwi Lim; Jeong, Taekyeong

    2016-01-01

    This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. The in-vessel STs of PGSFR will be estimated using the methodology of ANL-ART-38 report in additional to 4S methodology. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the realistic burnup conditions. ST in the release from the core to primary sodium is calculated by using the assumption of ANL methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using equation and experimental materials

  9. Development of design Criteria for ITER In-vessel Components

    International Nuclear Information System (INIS)

    Sannazzaro, G.; Barabash, V.; Kang, S.C.; Fernandez, E.; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, I.; Fernández, F.; Guirao, J.

    2013-01-01

    Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations

  10. 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)

  11. Estimation of the lifetime of resin insulators against baking temperature for JT-60SA in-vessel coils

    Energy Technology Data Exchange (ETDEWEB)

    Sukegawa, Atsuhiko M., E-mail: morioka.atsuhiko@jaea.go.jp; Murakami, Haruyuki; Matsunaga, Go; Sakurai, Shinji; Takechi, Manabu; Yoshida, Kiyoshi; Ikeda, Yoshitaka

    2015-10-15

    Highlights: • The lifetime of resin insulators at about 200 °C was estimated. • We make use of the Arrhenius plot by the Weibull analysis for the estimation. • A suitable temperatures for the in-vessel coils were discussed. - Abstract: In the present study, the thermal endurance of epoxy-based, bismaleimides, and cyanate ester resins for the current design of the in-vessel coils was measured by performing acceleration tests to assess their insulation properties using the thermal endurance defined by the International Electrotechnical Commission (IEC-60216 Part1–Part 6) for a minimum of 5,000 h in the 180–240 °C temperature range. It was found that none of the resin insulators could tolerate the baking conditions of 40,000 h at ∼200 °C in the JT-60SA vacuum vessel. Therefore, the design of the in-vessel coils, including the error field correction coils (EFCC), was changed from the type without water cooling to with water cooling on JT-60SA.

  12. Plasma arc melting of zirconium

    International Nuclear Information System (INIS)

    Tubesing, P.K.; Korzekwa, D.R.; Dunn, P.S.

    1997-01-01

    Zirconium, like some other refractory metals, has an undesirable sensitivity to interstitials such as oxygen. Traditionally, zirconium is processed by electron beam melting to maintain minimum interstitial contamination. Electron beam melted zirconium, however, does not respond positively to mechanical processing due to its large grain size. The authors undertook a study to determine if plasma arc melting (PAM) technology could be utilized to maintain low interstitial concentrations and improve the response of zirconium to subsequent mechanical processing. The PAM process enabled them to control and maintain low interstitial levels of oxygen and carbon, produce a more favorable grain structure, and with supplementary off-gassing, improve the response to mechanical forming

  13. Detonation waves in melt-coolant interaction. Part 2. Applied analysis

    International Nuclear Information System (INIS)

    Kolev, N.I.; Hulin, H.

    2001-01-01

    Making use of the detonation theory presented in part 1 for melt-water interaction, detonation solutions for different melt-water pairs at different conditions are compared to each other. Discussion is provided on the existence of detonation solutions for water droplet - melt droplet - gas systems. The conclusion is made that even if such solution can be realized in the nature, which is highly questionable, the resulting detonation pressures will be below 200 bar. This is an important result for judging the risk of the melt-water disperse mixtures in nuclear safety analysis. In addition, the detonation pressures for alumna-continuous water systems have been found to be stronger then those for urania-continuous water systems, in agreement with the experimental observations and seems to give finally the searched for a long time explanation why alumna-water systems detonate much more violent than urania-water systems. (orig.) [de

  14. Melt propagation in dry core debris beds

    International Nuclear Information System (INIS)

    Dosanjh, S.S.

    1989-01-01

    During severe light water reactor accidents like Three Mile Island Unit 2, the fuel rods can fragment and thus convert the reactor core into a large particle bed. The postdryout meltdown of such debris beds is examined. A two-dimensional model that considers the presence of oxidic (UO 2 and ZrO 2 ) as well as metallic (e.g., zirconium) constituents is developed. Key results are that a dense metallic crust is created near the bottom of the bed as molten materials flow downward and freeze; liquid accumulates above the blockage and, if zirconium is present, the pool grows rapidly as molten zirconium dissolved both UO 2 and ZrO 2 particles; if the melt wets the solid, a fraction of the melt flows radially outward under the action of capillary forces and freezes near the radial boundary; in a nonwetting system, all of the melt flows into the bottom of the bed; and when zirconium and iron are in intimate contact and the zirconium metal atomic fraction is > 0.33, these metals can liquefy and flow out of the bed very early in the meltdown sequence

  15. Nitrogen Control in VIM Melts

    Science.gov (United States)

    Jablonski, P. D.; Hawk, J. A.

    NETL has developed a design and control philosophy for the addition of nitrogen to austenitic and ferritic steels. The design approach uses CALPHAD as the centerpiece to predict the level to which nitrogen is soluble in both the melt and the solid. Applications of this technique have revealed regions of "exclusion" in which the alloy, while within specification limits of prescribed, cannot be made by conventional melt processing. Furthermore, other investigations have found that substantial retrograde solubility of nitrogen exists, which can become problematic during subsequent melt processing and/or other finishing operations such as welding. Additionally, the CALPHAD method has been used to adjust primary melt conditions. To that end, nitrogen additions have been made using chrome nitride, silicon nitride, high-nitrogen ferrochrome as well as nitrogen gas. The advantages and disadvantages of each approach will be discussed and NETL experience in this area will be summarized with respect to steel structure.

  16. Theoretical melting curve of caesium

    International Nuclear Information System (INIS)

    Simozar, S.; Girifalco, L.A.; Pennsylvania Univ., Philadelphia

    1983-01-01

    A statistical-mechanical model is developed to account for the complex melting curve of caesium. The model assumes the existence of three different species of caesium defined by three different electronic states. On the basis of this model, the free energy of melting and the melting curve are computed up to 60 kbar, using the solid-state data and the initial slope of the fusion curve as input parameters. The calculated phase diagram agrees with experiment to within the experimental error. Other thermodynamic properties including the entropy and volume of melting were also computed, and they agree with experiment. Since the theory requires only one adjustable constant, this is taken as strong evidence that the three-species model is satisfactory for caesium. (author)

  17. Melting curves of gammairradiated DNA

    International Nuclear Information System (INIS)

    Hofer, H.; Altmann, H.; Kehrer, M.

    1978-08-01

    Melting curves of gammairradiated DNA and data derived of them, are reported. The diminished stability is explained by basedestruction. DNA denatures completely at room temperature, if at least every fifth basepair is broken or weakened by irradiation. (author)

  18. In vessel retention for VVER 1000 - Experimental work

    International Nuclear Information System (INIS)

    Batek, D.

    2015-01-01

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

  19. Issues and strategies for DEMO in-vessel component integration

    International Nuclear Information System (INIS)

    Bachmann, C.; Arbeiter, F.; Boccaccini, L.V.; Coleman, M.; Federici, G.; Fischer, U.; Kemp, R.; Maviglia, F.; Mazzone, G.; Pereslavtsev, P.; Roccella, R.; Taylor, N.; Villari, R.; Villone, F.; Wenninger, R.; You, J.-H.

    2016-01-01

    In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

  20. Locking mechanism for in-vessel components of tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, S.; Shimizu, K.; Koizumi, K.; Tada, E.

    1992-01-01

    The locking and unlocking mechanism for in-vessel replaceable components such as blanket modules, is one of the most critical issues of the tokamak fusion reactor, since the sufficient stiffness against the enormous electromagnetic loads and the easy replaceability are required. In this paper, the authors decide that a caulking cotter joint is worth initiating the R and D from veiwpoints of an effective use of space, a replaceability, a removability of nuclear heating, and a reliability. In this approach, the cotter driving (thrusting and plucking) mechanism is a critical technology. A flexible tube concept has been developed as the driving mechanism, where the stroke and driving force are obtained by a fat shape by the hydraulic pressure. The original normal tube is subjected to the working percentage of more than several hundreds percentage (from thickness of 1.2 mm to 0.2 mm) for plastically forming the flexible tube

  1. Issues and strategies for DEMO in-vessel component integration

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, C., E-mail: christian.bachmann@euro-fusion.org [EUROfusion PMU, Garching (Germany); Arbeiter, F.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Coleman, M.; Federici, G. [EUROfusion PMU, Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kemp, R. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Maviglia, F. [EUROfusion PMU, Garching (Germany); Mazzone, G. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Roccella, R. [ITER Organization, St. Paul Lez Durance (France); Taylor, N. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Villari, R. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Villone, F. [ENEA-CREATE Association, DIEI, Università di Cassino e del Lazio Meridiona (Italy); Wenninger, R. [EUROfusion PMU, Garching (Germany); You, J.-H. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany)

    2016-11-15

    In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

  2. Manufacture of EAST VS In-Vessel Coil

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  3. Run-off of strontium with melting snow in spring

    International Nuclear Information System (INIS)

    Quenild, C.; Tveten, U.

    1986-09-01

    When assessing the consequences of atmospheric releases caused by a large reactor accident, one usually finds that the major contributions to the dose are via nutrition and from exposure to radiation from radioactive materials deposited on ground. The experiment described is concerned with run-off from agricultural surface which has been contaminated with strontiom while covered with snow. Migration experiments show a significant difference between summer and winter conditions. Roughly 54% of the strontium with which the experimental area was contaminated, ran off with the melt-water. Under winter conditions, portions of the contaminant will flow with the melt-water without coming in contact with the soil

  4. Melting in super-earths.

    Science.gov (United States)

    Stixrude, Lars

    2014-04-28

    We examine the possible extent of melting in rock-iron super-earths, focusing on those in the habitable zone. We consider the energetics of accretion and core formation, the timescale of cooling and its dependence on viscosity and partial melting, thermal regulation via the temperature dependence of viscosity, and the melting curves of rock and iron components at the ultra-high pressures characteristic of super-earths. We find that the efficiency of kinetic energy deposition during accretion increases with planetary mass; considering the likely role of giant impacts and core formation, we find that super-earths probably complete their accretionary phase in an entirely molten state. Considerations of thermal regulation lead us to propose model temperature profiles of super-earths that are controlled by silicate melting. We estimate melting curves of iron and rock components up to the extreme pressures characteristic of super-earth interiors based on existing experimental and ab initio results and scaling laws. We construct super-earth thermal models by solving the equations of mass conservation and hydrostatic equilibrium, together with equations of state of rock and iron components. We set the potential temperature at the core-mantle boundary and at the surface to the local silicate melting temperature. We find that ancient (∼4 Gyr) super-earths may be partially molten at the top and bottom of their mantles, and that mantle convection is sufficiently vigorous to sustain dynamo action over the whole range of super-earth masses.

  5. Pressure-induced melting of micellar crystal

    DEFF Research Database (Denmark)

    Mortensen, K.; Schwahn, D.; Janssen, S.

    1993-01-01

    that pressure improves the solvent quality of water, thus resulting in decomposition of the micelles and consequent melting of the micellar crystal. The combined pressure and temperature dependence reveals that in spite of the apparent increase of order on the 100 angstrom length scale upon increasing......Aqueous solutions of triblock copolymers of poly(ethylene oxide) and poly(propylene oxide) aggregate at elevated temperatures into micelles which for polymer concentrations greater-than-or-equal-to 20% make a hard sphere crystallization to a cubic micellar crystal. Structural studies show...... temperature (decreasing pressure) the overall entropy increases through the inverted micellar crystallization characteristic....

  6. Melting the vacuum

    International Nuclear Information System (INIS)

    Rafelski, J.

    1998-01-01

    Results presented at the Quark Matter 97 conference, held in December in Tsukuba, Japan, have provided new insights into the confinement of quarks in matter. The current physics paradigm is that the inertial masses of protons and neutrons, and hence of practically all of the matter around us, originate in the zero-point energy caused by the confinement of quarks inside the small volume of the nucleon. Today, 25 years after Harald Fritzsch, Heinrich Leutwyler and Murray Gell-Mann proposed quantum chromodynamics (QCD) as a means for understanding strongly interacting particles such as nucleons and mesons, our understanding of strong interactions and quark confinement remains incomplete. Quarks and the gluons that bind them together have a ''colour'' charge that may be red, green or blue. But quarks are seen in particles that are white: baryons such as protons and neutrons consist of three quarks with different colour charges, while mesons consist of a quark and an antiquark, and again the colour charge cancels out. To prove that confinement arises from quark-gluon fluctuations in the vacuum that quantum theories dictate exists today, we need to find a way of freeing the colour charge of quarks. Experiments must therefore ''melt'' the vacuum to deconfine quarks and the colour charge. By colliding nuclei at high energies, we hope to produce regions of space filled with free quarks and gluons. This deconfined phase is known as the quark-gluon plasma. At the Tsukuba meeting, Scott Pratt of Michigan State University in the US discussed measurements that show that the hot dense state of matter created in these collisions exists for only 2x10 -23 s. So does the quark gluon plasma exist? No-one doubts that it did at one time, before the vacuum froze into its current state about 20 into the life of the universe, causing the nucleons to form as we know them today. The issue is whether we can recreate this early stage of the universe in laboratory experiments. And if we did

  7. The refreezing of melt ponds on Arctic sea ice

    Science.gov (United States)

    Flocco, Daniela; Feltham, Daniel L.; Bailey, Eleanor; Schroeder, David

    2015-02-01

    The presence of melt ponds on the surface of Arctic sea ice significantly reduces its albedo, inducing a positive feedback leading to sea ice thinning. While the role of melt ponds in enhancing the summer melt of sea ice is well known, their impact on suppressing winter freezing of sea ice has, hitherto, received less attention. Melt ponds freeze by forming an ice lid at the upper surface, which insulates them from the atmosphere and traps pond water between the underlying sea ice and the ice lid. The pond water is a store of latent heat, which is released during refreezing. Until a pond freezes completely, there can be minimal ice growth at the base of the underlying sea ice. In this work, we present a model of the refreezing of a melt pond that includes the heat and salt balances in the ice lid, trapped pond, and underlying sea ice. The model uses a two-stream radiation model to account for radiative scattering at phase boundaries. Simulations and related sensitivity studies suggest that trapped pond water may survive for over a month. We focus on the role that pond salinity has on delaying the refreezing process and retarding basal sea ice growth. We estimate that for a typical sea ice pond coverage in autumn, excluding the impact of trapped ponds in models overestimates ice growth by up to 265 million km3, an overestimate of 26%.

  8. Research of Snow-Melt Process on a Heated Platform

    Directory of Open Access Journals (Sweden)

    Vasilyev Gregory P.

    2016-01-01

    Full Text Available The article has shown the results of experimental researches of the snow-melt on a heated platform-near building heat-pump snow-melt platform. The near-building (yard heat pump platforms for snow melt with the area up to 10-15 m2 are a basis of the new ideology of organization of the street cleaning of Moscow from snow in the winter period which supposes the creation in the megalopolis of the «distributed snow-melt system» (DSMS using non-traditional energy sources. The results of natural experimental researches are presented for the estimation of efficiency of application in the climatic conditions of Moscow of heat pumps in the snow-melt systems. The researches were conducted on a model sample of the near-building heat-pump platform which uses the low-potential thermal energy of atmospheric air. The conducted researches have confirmed experimentally in the natural conditions the possibility and efficiency of using of atmospheric air as a source of low-potential thermal energy for evaporation of the snow-melt heat pump systems in the climatic conditions of Moscow. The results of laboratory researches of snow-melt process on a heated horizontal platform are presented. The researches have revealed a considerable dependence of efficiency of the snow-melt process on its piling mode (form-building and the organization of the process of its piling mode (form-building and the organization of the process of its (snow mass heat exchange with the surface of the heated platform. In the process of researches the effect of formation of an «ice dome» under the melting snow mass called by the fact that in case of the thickness of snow loaded on the platform more than 10 cm the water formed from the melting snow while the contact with the heating surface don’t spread on it, but soaks into the snow, wets it due to capillary effect and freezes. The formation of «ice dome» leads to a sharp increase of snow-melt period and decreases the operating

  9. Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Li, C., E-mail: chuanfei.li@ipp.mpg.de; Boscary, J.; Dekorsy, N.; Junghanns, P.; Mendelevitch, B.; Peacock, A.; Pirsch, H.; Sellmeier, O.; Springer, J.; Stadler, R.; Streibl, B.

    2014-10-15

    Highlights: • Thousand parts for the divertor, first wall, cooling supply and diagnostics as W7-X In-Vessel Components. • Database building including part and assembly data, work and capacity organization, quality assurance documents. • Production management system to organize the fabrication and the associated quality assurance. • Successful use of an efficient and flexible product planning and scheduling tool for W7-X In-Vessel Components. - Abstract: The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements.

  10. Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

    International Nuclear Information System (INIS)

    Li, C.; Boscary, J.; Dekorsy, N.; Junghanns, P.; Mendelevitch, B.; Peacock, A.; Pirsch, H.; Sellmeier, O.; Springer, J.; Stadler, R.; Streibl, B.

    2014-01-01

    Highlights: • Thousand parts for the divertor, first wall, cooling supply and diagnostics as W7-X In-Vessel Components. • Database building including part and assembly data, work and capacity organization, quality assurance documents. • Production management system to organize the fabrication and the associated quality assurance. • Successful use of an efficient and flexible product planning and scheduling tool for W7-X In-Vessel Components. - Abstract: The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements

  11. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  12. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  13. Sulfur concentration at sulfide saturation (SCSS) in magmatic silicate melts

    Science.gov (United States)

    Liu, Yanan; Samaha, Naji-Tom; Baker, Don R.

    2007-04-01

    The sulfur concentration in silicate melts at sulfide saturation (SCSS) was experimentally investigated in a temperature range from 1150 to 1450 °C and a pressure range from 500 MPa to 1 GPa in a piston-cylinder apparatus. The investigated melt compositions varied from rhyolitic to basaltic and water concentrations varied from 0 to ˜9 wt%. All experiments were saturated with FeS melt or pyrrhotite crystals. Temperature was confirmed to have a positive effect on the SCSS. Experimental oxygen fugacities were either near the carbon-carbon monoxide buffer or one log unit above the nickel-nickel oxide buffer, and found to positively affect the SCSS. Combining our results with data from the literature we constructed a model to predict the SCSS in melts ranging in composition from komatiitic to rhyolitic, with water concentrations from 0 to 9 wt%, at pressures from 1 bar to 9 GPa and oxygen fugacities between ˜2 log units below the fayalite-magnetite-quartz buffer to ˜2 log units above it. The coefficients were obtained by multiple linear regression of experimental data and the best model found for the prediction of the SCSS is: ln(Sinppm)=11.35251-{4454.6}/{T}-0.03190{P}/{T}+0.71006ln(MFM)-1.98063[(MFM)(XO)]+0.21867ln(XO)+0.36192lnX where P is in bar, T is in K, MFM is a compositional parameter describing the melt based upon cation mole fractions: MFM={Na+K+2(Ca+Mg+Fe)}/{Si×(Al+Fe)}, XO is the mole fraction of water in the melt, and X is the mole fraction of FeO in the melt. This model was independently tested against experiments performed on anhydrous and hydrous melts in the temperature range from 800 to 1800 °C and 1-9 GPa. The model typically predicts the measured values of the natural log of the SCSS (in ppm) for komatiitic to rhyolitic (˜42 to ˜74 wt% SiO 2) melts to within 5% relative, but is less accurate for high-silica (>76 wt% SiO 2) rhyolites, especially those with molar ratios of iron to sulfur below 2. We demonstrate how this model can be used with

  14. Methods for Melting Temperature Calculation

    Science.gov (United States)

    Hong, Qi-Jun

    Melting temperature calculation has important applications in the theoretical study of phase diagrams and computational materials screenings. In this thesis, we present two new methods, i.e., the improved Widom's particle insertion method and the small-cell coexistence method, which we developed in order to capture melting temperatures both accurately and quickly. We propose a scheme that drastically improves the efficiency of Widom's particle insertion method by efficiently sampling cavities while calculating the integrals providing the chemical potentials of a physical system. This idea enables us to calculate chemical potentials of liquids directly from first-principles without the help of any reference system, which is necessary in the commonly used thermodynamic integration method. As an example, we apply our scheme, combined with the density functional formalism, to the calculation of the chemical potential of liquid copper. The calculated chemical potential is further used to locate the melting temperature. The calculated results closely agree with experiments. We propose the small-cell coexistence method based on the statistical analysis of small-size coexistence MD simulations. It eliminates the risk of a metastable superheated solid in the fast-heating method, while also significantly reducing the computer cost relative to the traditional large-scale coexistence method. Using empirical potentials, we validate the method and systematically study the finite-size effect on the calculated melting points. The method converges to the exact result in the limit of a large system size. An accuracy within 100 K in melting temperature is usually achieved when the simulation contains more than 100 atoms. DFT examples of Tantalum, high-pressure Sodium, and ionic material NaCl are shown to demonstrate the accuracy and flexibility of the method in its practical applications. The method serves as a promising approach for large-scale automated material screening in which

  15. A metastable liquid melted from a crystalline solid under decompression

    Science.gov (United States)

    Lin, Chuanlong; Smith, Jesse S.; Sinogeikin, Stanislav V.; Kono, Yoshio; Park, Changyong; Kenney-Benson, Curtis; Shen, Guoyin

    2017-01-01

    A metastable liquid may exist under supercooling, sustaining the liquid below the melting point such as supercooled water and silicon. It may also exist as a transient state in solid-solid transitions, as demonstrated in recent studies of colloidal particles and glass-forming metallic systems. One important question is whether a crystalline solid may directly melt into a sustainable metastable liquid. By thermal heating, a crystalline solid will always melt into a liquid above the melting point. Here we report that a high-pressure crystalline phase of bismuth can melt into a metastable liquid below the melting line through a decompression process. The decompression-induced metastable liquid can be maintained for hours in static conditions, and transform to crystalline phases when external perturbations, such as heating and cooling, are applied. It occurs in the pressure-temperature region similar to where the supercooled liquid Bi is observed. Akin to supercooled liquid, the pressure-induced metastable liquid may be more ubiquitous than we thought.

  16. Melting of superheated molecular crystals

    Science.gov (United States)

    Cubeta, Ulyana; Bhattacharya, Deepanjan; Sadtchenko, Vlad

    2017-07-01

    Melting dynamics of micrometer scale, polycrystalline samples of isobutane, dimethyl ether, methyl benzene, and 2-propanol were investigated by fast scanning calorimetry. When films are superheated with rates in excess of 105 K s-1, the melting process follows zero-order, Arrhenius-like kinetics until approximately half of the sample has transformed. Such kinetics strongly imply that melting progresses into the bulk via a rapidly moving solid-liquid interface that is likely to originate at the sample's surface. Remarkably, the apparent activation energies for the phase transformation are large; all exceed the enthalpy of vaporization of each compound and some exceed it by an order of magnitude. In fact, we find that the crystalline melting kinetics are comparable to the kinetics of dielectric α-relaxation in deeply supercooled liquids. Based on these observations, we conclude that the rate of non-isothermal melting for superheated, low-molecular-weight crystals is limited by constituent diffusion into an abnormally dense, glass-like, non-crystalline phase.

  17. Improved capacitive melting curve measurements

    International Nuclear Information System (INIS)

    Sebedash, Alexander; Tuoriniemi, Juha; Pentti, Elias; Salmela, Anssi

    2009-01-01

    Sensitivity of the capacitive method for determining the melting pressure of helium can be enhanced by loading the empty side of the capacitor with helium at a pressure nearly equal to that desired to be measured and by using a relatively thin and flexible membrane in between. This way one can achieve a nanobar resolution at the level of 30 bar, which is two orders of magnitude better than that of the best gauges with vacuum reference. This extends the applicability of melting curve thermometry to lower temperatures and would allow detecting tiny anomalies in the melting pressure, which must be associated with any phenomena contributing to the entropy of the liquid or solid phases. We demonstrated this principle in measurements of the crystallization pressure of isotopic helium mixtures at millikelvin temperatures by using partly solid pure 4 He as the reference substance providing the best possible universal reference pressure. The achieved sensitivity was good enough for melting curve thermometry on mixtures down to 100 μK. Similar system can be used on pure isotopes by virtue of a blocked capillary giving a stable reference condition with liquid slightly below the melting pressure in the reference volume. This was tested with pure 4 He at temperatures 0.08-0.3 K. To avoid spurious heating effects, one must carefully choose and arrange any dielectric materials close to the active capacitor. We observed some 100 pW loading at moderate excitation voltages.

  18. Automatic Control of Silicon Melt Level

    Science.gov (United States)

    Duncan, C. S.; Stickel, W. B.

    1982-01-01

    A new circuit, when combined with melt-replenishment system and melt level sensor, offers continuous closed-loop automatic control of melt-level during web growth. Installed on silicon-web furnace, circuit controls melt-level to within 0.1 mm for as long as 8 hours. Circuit affords greater area growth rate and higher web quality, automatic melt-level control also allows semiautomatic growth of web over long periods which can greatly reduce costs.

  19. Proton NMR relaxation in hydrous melts

    International Nuclear Information System (INIS)

    Braunstein, J.; Bacarella, A.L.; Benjamin, B.M.; Brown, L.L.; Girard, C.

    1976-01-01

    Pulse and continuous wave NMR measurements are reported for protons in hydrous melts of calcium nitrate at temperatures between -4 and 120 0 C. Although measured in different temperature ranges, spin-lattice (T 1 ) and spin-spin (T 2 ) relaxation times appear to be nearly equal to each other and proportional to the self-diffusion coefficients of solute metal cations such as Cd 2+ . At temperatures near 50 0 C, mean Arrhenius coefficients Δ H/sub T 1 / (kcal/mol) are 7.9, 7.3, and 4.8, respectively, for melts containing 2.8, 4.0, and 8.0 moles of water per mole of calcium nitrate, compared to 4.6 kcal/mol for pure water. Temperature dependence of T 1 and T 2 in Ca(NO 3 ) 2 -2.8 H 2 O between -4 and 120 0 C are non-Arrhenius and can be represented by a Fulcher-type equation with a ''zero mobility temperature'' (T 0 ) of 225 0 K, close to the value of T 0 for solute diffusion, electrical conductance and viscosity. Resolution of the relaxation rates into correlation times for intramolecular (rotational) and intermolecular (translational) diffusional motion is discussed in terms of the Bloembergen-Purcell-Pound and more recent models for dipolar relaxation

  20. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

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

    1997-01-01

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  1. Magnetic susceptibility of semiconductor melts

    International Nuclear Information System (INIS)

    Kutvitskij, V.A.; Shurygin, P.M.

    1975-01-01

    The temperature dependences chi of various alloys confirm the existence of cluster formations in molten semiconductors, the stability of these formations in melts being considerably affected by the anion nature. The concentrational dependences of the magnetic susceptibility for all the investigated systems exhibit the diamagnetism maxima corresponding to the compound compositions. Heating the melt causes ''smearing'' the maxima, which is related with the cluster structure dissociation. The existence of the maxima concentrational dependence chi corresponding to BiTe and BiSe is found in the isotherms. The non-linear dependence of chi on the composition shows the absence of a single-valued relation between the phase diagram and the chi-diagram for melts

  2. 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

  3. Crust behavior and erosion rate prediction of EPR sacrificial material impinged by core melt jet

    Energy Technology Data Exchange (ETDEWEB)

    Li, Gen; Liu, Ming, E-mail: ming.liu@mail.xjtu.edu.cn; Wang, Jinshi; Chong, Daotong; Yan, Junjie

    2017-04-01

    Highlights: • A numerical code was developed to analyze melt jet-concrete interaction in the frame of MPS method. • Crust and ablated concrete layer at UO{sub 2}-ZrO{sub 2} melt and concrete interface periodically developed and collapsed. • Concrete surface temperature fluctuated around a low temperature and ablation temperature. • Concrete erosion by Fe-Zr melt jet was significantly faster than that by UO{sub 2}-ZrO{sub 2} melt jet. - Abstract: Sacrificial material is a special ferro-siliceous concrete, designed in the ex-vessel core melt stabilization system of European Pressurized water Reactor (EPR). Given a localized break of RPV lower head, the melt directly impinges onto the dry concrete in form of compact jet. The concrete erosion behavior influences the failure of melt plug, and further affects melt spreading. In this study, a numerical code was developed in the frame of Moving Particle Semi-implicit (MPS) method, to analyze the crust behavior and erosion rate of sacrificial concrete, impinged by prototypic melt jet. In validation of numerical modeling, the time-dependent erosion depth and erosion configuration matched well with the experimental data. Sensitivity study of sacrificial concrete erosion indicates that the crust and ablated concrete layer presented at UO{sub 2}-ZrO{sub 2} melt and concrete interface, whereas no crust could be found in the interaction of Fe-Zr melt with concrete. The crust went through stabilization-fracture-reformation periodic process, accompanied with accumulating and collapsing of molten concrete layer. The concrete surface temperature fluctuated around a low temperature and ablation temperature. It increased as the concrete surface layer was heated to melting, and dropped down when the cold concrete was revealed. The erosion progression was fast in the conditions of small jet diameter and large concrete inclination angle, and it was significantly faster in the erosion by metallic melt jet than by oxidic melt jet.

  4. Sensitivity analyses on in-vessel hydrogen generation for KNGR

    International Nuclear Information System (INIS)

    Kim, See Darl; Park, S.Y.; Park, S.H.; Park, J.H.

    2001-03-01

    Sensitivity analyses for the in-vessel hydrogen generation, using the MELCOR program, are described in this report for the Korean Next Generation Reactor. The typical accident sequences of a station blackout and a large LOCA scenario are selected. A lower head failure model, a Zircaloy oxidation reaction model and a B 4 C reaction model are considered for the sensitivity parameters. As for the base case, 1273.15K for a failure temperature of the penetrations or the lower head, an Urbanic-Heidrich correlation for the Zircaloy oxidation reaction model and the B 4 C reaction model are used. Case 1 used 1650K as the failure temperature for the penetrations and Case 2 considered creep rupture instead of penetration failure. Case 3 used a MATPRO-EG and G correlation for the Zircaloy oxidation reaction model and Case 4 turned off the B 4 C reaction model. The results of the studies are summarized below : (1) When the penetration failure temperature is higher, or the creep rupture failure model is considered, the amount of hydrogen increases for two sequences. (2) When the MATPRO-EG and G correlation for a Zircaloy oxidation reaction is considered, the amount of hydrogen is less than the Urbanic-Heidrich correlation (Base case) for both scenarios. (3) When the B 4 C reaction model turns off, the amount of hydrogen decreases for two sequences

  5. R and D on ITER in-vessel magnetic sensors

    International Nuclear Information System (INIS)

    Peruzzo, Simone; Arshad, Shakeib; Brombin, Matteo; Chitarin, Giuseppe; Gonzalez, Winder; Grando, Luca; Portales, Mickael; Rizzolo, Andrea; Vayakis, George; Vermeeren, Ludo

    2013-01-01

    Highlights: ► Magnetic sensors based on LTCC technology should be able to meet ITER requirements. ► Outgassing, TIEMF and RIEMF tests are planned on new prototypes. ► The design of a proper sensor support and connection system is in progress. -- Abstract: This paper summarizes the progress of R and D activities related to the design of in-vessel magnetic sensors for measurement of the equilibrium field in ITER and the associated mechanical support and connection system. The background to the design activities, developed in the last few years in the framework of several collaborations, is provided in the introduction. The paper focuses on the experimental results obtained from a set of sensor prototypes, recently manufactured and tested, and on numerical simulations performed in order to verify the compliance of the sensors with ITER requirements. The paper finally illustrates the status of the engineering analyses performed to progress the design of the support and connection system, conceived to be replaceable by remote handling. In conclusion, specific issues which require further developments to achieve the final design are outlined

  6. In-vessel dust and tritium control strategy in ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

  7. Conceptual design of EAST flexible in-vessel inspection system

    International Nuclear Information System (INIS)

    Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

    2010-01-01

    Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

  8. R and D on ITER in-vessel magnetic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Peruzzo, Simone [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Arshad, Shakeib [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Brombin, Matteo, E-mail: matteo.brombin@igi.cnr.it [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Chitarin, Giuseppe; Gonzalez, Winder; Grando, Luca [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Portales, Mickael [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Rizzolo, Andrea [Consorzio RFX, Association EURATOM-ENEA, C.so Stati Uniti 4, 35127 Padova (Italy); Vayakis, George [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Vermeeren, Ludo [SCK-CEN, Institute of Advanced Nuclear Systems, Boeretang 200, B-2400 Mol (Belgium)

    2013-10-15

    Highlights: ► Magnetic sensors based on LTCC technology should be able to meet ITER requirements. ► Outgassing, TIEMF and RIEMF tests are planned on new prototypes. ► The design of a proper sensor support and connection system is in progress. -- Abstract: This paper summarizes the progress of R and D activities related to the design of in-vessel magnetic sensors for measurement of the equilibrium field in ITER and the associated mechanical support and connection system. The background to the design activities, developed in the last few years in the framework of several collaborations, is provided in the introduction. The paper focuses on the experimental results obtained from a set of sensor prototypes, recently manufactured and tested, and on numerical simulations performed in order to verify the compliance of the sensors with ITER requirements. The paper finally illustrates the status of the engineering analyses performed to progress the design of the support and connection system, conceived to be replaceable by remote handling. In conclusion, specific issues which require further developments to achieve the final design are outlined.

  9. Review of SFR In-Vessel Radiological Source Term Studies

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2008-10-01

    An effort has been made in this study to search for and review the literatures in public domain on the studies of the phenomena related to the release of radionuclides and aerosols to the reactor containment of the sodium fast reactor (SFR) plants (i.e., in-vessel source term), made in Japan and Europe including France, Germany and UK over the last few decades. Review work is focused on the experimental programs to investigate the phenomena related to determining the source terms, with a brief review on supporting analytical models and computer programs. In this report, the research programs conducted to investigate the CDA (core disruptive accident) bubble behavior in the sodium pool for determining 'primary' or 'instantaneous' source term are first introduced. The studies performed to determine 'delayed source term' are then described, including the various stages of phenomena and processes: fission product (FP) release from fuel , evaporation release from the surface of the pool, iodine mass transfer from fission gas bubble, FP deposition , and aerosol release from core-concrete interaction. The research programs to investigate the release and transport of FPs and aerosols in the reactor containment (i.e., in-containment source term) are not described in this report

  10. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  11. Dynamic simulation of the NET In-Vessel Handling Unit

    International Nuclear Information System (INIS)

    Reim, J.

    1991-01-01

    During the conceptual design phase of the Next European Torus (NET) a large remote maintenance transporter system, the In-Vessel Handling Unit (IVHU), is being developed. It consists of an articulated boom with four rotational joints, which is mounted on a carrier outside the vessel. This boom will be able to carry master-slave manipulators or special work units. The engineering design is supported by dynamic computations. Main topics of the dynamic simulation are the evaluation of IVHU performance, selection and optimisation of the actuator design and of the control algorithms. This simulation task requires full three-dimensional modelling regarding structural elasticity and non-linear actuator dynamics. The Multibody dynamics of the transporter system are modelled with a commerical analysis package. Elastic links and a precise dynamic actuator model are introduced by applied forces, spring elements and differential equations. The actuator model comprises electric motors, gears and linear control algorithms. Non-linear effects which have an influence on control stability and accuracy are taken into account. Most important effects are backlash and static friction. The simulations concentrate on test and optimisation of the control layout and performance studies for critical remote handling tasks. Simulations for control layout and critical remote maintenance tasks correspond to the design objectives of the transporter system. (orig.)

  12. Melt-assisted synthesis to lanthanum hexaboride nanoparticles and ...

    Indian Academy of Sciences (India)

    2017-09-22

    Sep 22, 2017 ... cubes (100–300 nm) when metal zinc was used as reaction melt. Photothermal ... powder (Zn, AR), potassium chloride (KCl, AR), sodium chloride (NaCl, AR) and ... ized water to remove NaCl, KCl and other impurities. The.

  13. On the rapid melt quenching

    International Nuclear Information System (INIS)

    Usatyuk, I.I.; Novokhatskij, I.A.; Kaverin, Yu.F.

    1994-01-01

    Specific features of instrumentation of traditionally employed method of melt spinning (rapid quenching), its disadvantages being discussed, were analyzed. The necessity of the method upgrading as applied to the problems of studying fine structure of molten metals and glasses was substantiated. The principle flowsheet of experimental facility for extremely rapid quenching of the melts of metals is described, specificity of its original functional units being considered. The sequence and character of all the principal stages of the method developed were discussed. 18 refs.; 3 figs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  15. Experiments on melt dispersion with lateral failure in the bottom head of the pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, L.; Gargallo, M. [Forschungszentrum Karlsruhe, Institut fur Kern-und Energietechnik, Karlsruhe (Germany)

    2001-07-01

    Melt dispersion experiments with lateral failure in the bottom head were carried out in a 1:18 scaled annular cavity design under low pressure conditions. Water and a bismuth alloy were used as melt simulant material and nitrogen as driving gas. With lateral breaches the liquid height in the lower head relative to the upper and lower edge of the breach is an additional parameter for the dispersion process. Shifting the break from the central position towards the side of the lower head leads to smaller melt dispersion, and a larger breach size does not necessarily lead to a larger dispersed melt fraction. (author)

  16. Additive manufacturing of ITER first wall panel parts by two approaches: Selective laser melting and electron beam melting

    International Nuclear Information System (INIS)

    Zhong, Yuan; Rännar, Lars-Erik; Wikman, Stefan; Koptyug, Andrey; Liu, Leifeng; Cui, Daqing; Shen, Zhijian

    2017-01-01

    Highlights: • A novel way using additive manufacturing to fabricated ITER First Wall Panel parts is proposed. • ITER First Wall Panel parts successfully manufactured by both SLM and EBM are compared. • Physical and mechanical properties of SLM and EBM SS316L are clearly compared. • Problems encountered for large scale part building were discussed and possible solutions are given. - Abstract: Fabrication of ITER First Wall (FW) Panel parts by two additive manufacturing (AM) technologies, selective laser melting (SLM) and electron beam melting (EBM), was supported by Fusion for Energy (F4E). For the first time, AM is applied to manufacture ITER In-Vessel parts with complex design. Fully dense SS316L was prepared by both SLM and EBM after developing optimized laser/electron beam parameters. Characterizations on the density, magnetic permeability, microstructure, defects and inclusions were carried out. Tensile properties, Charpy-impact properties and fatigue properties of SLM and EBM SS316L were also compared. ITER FW Panel parts were successfully fabricated by both SLM and EBM in a one-step building process. The SLM part has smoother surface, better size accuracy while the EBM part takes much less time to build. Issues with removing support structures might be solved by slightly changing the design of the internal cooling system. Further investigation of the influence of neutron irradiation on materials properties between the two AM technologies is needed.

  17. Additive manufacturing of ITER first wall panel parts by two approaches: Selective laser melting and electron beam melting

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Yuan [Department of Materials and Environmental Chemistry, Arrhenius Laboratory, Stockholm University, SE-106 91 Stockholm (Sweden); Rännar, Lars-Erik [Department of Quality Technology, Mechanical Engineering and Mathematics, Sports Tech Research Centre, Mid Sweden University, SE-831 25 Östersund (Sweden); Wikman, Stefan [Fusion for Energy, Torres Diagonal Litoral B3, Josep Pla 2, 08019 Barcelona (Spain); Koptyug, Andrey [Department of Quality Technology, Mechanical Engineering and Mathematics, Sports Tech Research Centre, Mid Sweden University, SE-831 25 Östersund (Sweden); Liu, Leifeng; Cui, Daqing [Department of Materials and Environmental Chemistry, Arrhenius Laboratory, Stockholm University, SE-106 91 Stockholm (Sweden); Shen, Zhijian, E-mail: shen@mmk.su.se [Department of Materials and Environmental Chemistry, Arrhenius Laboratory, Stockholm University, SE-106 91 Stockholm (Sweden)

    2017-03-15

    Highlights: • A novel way using additive manufacturing to fabricated ITER First Wall Panel parts is proposed. • ITER First Wall Panel parts successfully manufactured by both SLM and EBM are compared. • Physical and mechanical properties of SLM and EBM SS316L are clearly compared. • Problems encountered for large scale part building were discussed and possible solutions are given. - Abstract: Fabrication of ITER First Wall (FW) Panel parts by two additive manufacturing (AM) technologies, selective laser melting (SLM) and electron beam melting (EBM), was supported by Fusion for Energy (F4E). For the first time, AM is applied to manufacture ITER In-Vessel parts with complex design. Fully dense SS316L was prepared by both SLM and EBM after developing optimized laser/electron beam parameters. Characterizations on the density, magnetic permeability, microstructure, defects and inclusions were carried out. Tensile properties, Charpy-impact properties and fatigue properties of SLM and EBM SS316L were also compared. ITER FW Panel parts were successfully fabricated by both SLM and EBM in a one-step building process. The SLM part has smoother surface, better size accuracy while the EBM part takes much less time to build. Issues with removing support structures might be solved by slightly changing the design of the internal cooling system. Further investigation of the influence of neutron irradiation on materials properties between the two AM technologies is needed.

  18. Radiation balances of melting snow covers at an open site in the Central Sierra Nevada, California

    International Nuclear Information System (INIS)

    Aguado, E.

    1985-01-01

    The radiation balances of melting snow packs for three seasons at an open site at the Central Sierra Snow Laboratory near Soda Springs, California were examined. The snow covers were examples of below-normal, near-normal and much-above-normal water equivalents. Two of the snow covers melted under generally clear skies in late spring while the other melted under cloudier conditions and at a time when less extraterrestrial radiation was available. Moreover, the snow covers were of very different densities, thereby allowing examination of a possible relationship between that characteristic and albedo. No such relationship was observed. Despite the dissimilarities in the conditions under which melt occurred, the disposition of solar radiation was similar for the three melt seasons. Albedos and their rates of decline through the melt season were similar for the three seasons. Absorbed solar radiation and a cloudiness index were useful predictors for daily net radiation, accounting for 71% of the total variance. (author)

  19. Melting of Uranium Metal Powders with Residual Salts

    International Nuclear Information System (INIS)

    Jin-Mok Hur; Dae-Seung Kang; Chung-Seok Seo

    2007-01-01

    The Advanced Spent Fuel Conditioning Process (ACP) of the Korea Atomic Energy Research Institute focuses on the conditioning of Pressurized Water Reactor spent oxide nuclear fuel. After the oxide reduction step of the ACP, the resultant metal powders containing ∼ 30 wt% residual LiCl-Li 2 O should be melted for a consolidation of the fine metal powders. In this study, we investigated the melting behaviors of uranium metal powders considering the effects of a LiCl-Li 2 O residual salt. (authors)

  20. Acoustic detection of melt particles

    International Nuclear Information System (INIS)

    Costley, R.D. Jr.

    1988-01-01

    The Reactor Safety Research Department at Sandia National Laboratories is investigating a type of Loss of Coolant Accident (LOCA). In this particular type of accident, core meltdown occurs while the pressure within the reactor pressure vessel (RPV) is high. If one of the instrument tube penetrations in the lower head fails, melt particles stream through the cavity and into the containment vessel. This experiment, which simulates this type accident, was performed in the Surtsev Direct Heating Test Facility which is approximately a 1:10 linear scaling of a large dry containment volume. A 1:10 linear scale model of the reactor cavity was placed near the bottom of the Surtsey vessel so that the exit of the cavity was at the vertical centerline of the vessel. A pressure vessel used to create the simulated molten core debris was located at the scaled height of the RPV. In order to better understand how the melt leaves the cavity and streams into the containment an array of five acoustic sensors was placed directly in the path of the melt particles about 30 feet from the exit of the sealed cavity. Highly damped, broadband sensors were chosen to minimize ringing so that individual particle hits could be detected. The goal was to count the signals produced by the individual particle hits to get some idea of how the melt particles left the cavity. This document presents some of the results of the experiment. 9 figs

  1. Thermodynamics of freezing and melting

    DEFF Research Database (Denmark)

    Pedersen, Ulf Rørbæk; Costigliola, Lorenzo; Bailey, Nicholas

    2016-01-01

    phases at a single thermodynamic state point provide the basis for calculating the pressure, density and entropy of fusion as functions of temperature along the melting line, as well as the variation along this line of the reduced crystalline vibrational mean-square displacement (the Lindemann ratio...

  2. Isotopic fractionation and profile evolution of a melting snowcover

    Institute of Scientific and Technical Information of China (English)

    周石硚; 中尾正义; 桥本重将; 坂井亚规子; 成田英器; 石川信敬

    2001-01-01

    Successive snow pits were dug intensively in a melting snowcover. Water was successfully separated from snow grains in the field for the first time. By measuring δ18O values of water and snow grain samples as well as comparing isotopic profiles, it is found that meltwater percolating down in snow develops quick and clear isotopic fractionation with snow grains, but exerts no clear impact on the δ18O profile of the snowcover through which the meltwater percolates.

  3. Water

    Science.gov (United States)

    ... www.girlshealth.gov/ Home Nutrition Nutrition basics Water Water Did you know that water makes up more ... to drink more water Other drinks How much water do you need? top Water is very important, ...

  4. Melting of fuel element racks and their recycling as granulate

    International Nuclear Information System (INIS)

    Quade, U.; Kluth, T.; Kreh, R.

    1998-01-01

    In order to increase the storage capacity for spent fuel elements in the Spanish NPPs of Almaraz and Asco, the existing racks were replaced by compact one in 1991/1993. The 28 racks from Almaraz NPP were cut on site, packed in 200-I-drums and taken to intermediate storage. For the remaining 28 racks of Asco NPP, ENRESA preferred the melting alternative. To demonstrate the recycling path melting in Germany, a test campaign with six racks was performed in 1997. As a result of this test melt, the limits for Carla melting plant were modified to 200 Bq/g total, α, β, γ 100 Bq/g nuclear fuels, max. 3g/100 kg 2,000 Bq/g total Fe55, H 3 , C-14 and Ni63. After the test melt campaign, the German authorities licensed the import and treatment of the remaining 22 racks on the condition that the waste resulting from the melting process as well as the granules produced were taken back to Spain. The shipment from Asco via France to Germany has been carried out in F 20-ft-IPII containers in accordance with ADR. Size reduction to chargeable dimensions was carried out by a plasma burner and hydraulic shears. For melting, a 3.2 Mg medium frequency induction furnace, operated in a separate housing, was used. For granules production outside this housing, the liquid iron was cast into a 5Mg ladle and then, through a water jet, into the granulating basin. The total mass of 287,659 Kg of 28 fuel elements racks and components of the storage basin yielded 297,914 kg of iron granulate. Secondary waste from melting amounted to 9,920 kg, corresponding to 3.45% of the input mass. The granulating process produced 6,589 kg, corresponding to 2.28% of the total mass to be melted. Radiological analysis of samples taken from the melt and different waste components confirmed the main nuclides Co60, Cs134 and Cs137. Fe55 was highly overestimated by the preliminary analysis. (Author) 2 refs

  5. 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.

  6. Cold-crucible melting of hulls and structural materials

    International Nuclear Information System (INIS)

    Jouan, A.; Jacquet-Francillon, N.; Puyou, M.; Piccinato, R.

    1990-01-01

    The method currently implemented at the La Hague UP3 reprocessing plant for conditioning of PWR zircaloy hulls is cement embedding. Another promising method, mainly for reducing the waste volume and the available exchange surface area, is melting. A cold-crucible melting process has therefore been developed by the CEA at Marcoule (France) over the last decade. Development work first concentrated on cladding hulls from fast breeder reactors, then from pressurized water reactors. The process can be used for both types of cladding wastes. Subassembly head and foot end-caps are sheared off and should be suitable for surface storage after α decontamination by successive rinsing. If necessary because of their α activity, they could be melted in a larger furnace

  7. THE PHYSICS OF MELTING IN EARLY MODERN LOVE POETRY

    Directory of Open Access Journals (Sweden)

    Andrea Brady

    2014-12-01

    Full Text Available Melting is a familiar trope in early modern erotic poetry, where it can signify the desire to transform the beloved from icy chastity through the warmth of the lover’s passion. However, this Petrarchan convention can be defamiliarised by thinking about the experiences of freezing and melting in this period. Examining melting in the discourses of early modern meteorology, medicine, proverb, scientific experiments, and preservative technologies, as well as weather of the Little Ice Age and the exploration of frozen hinterlands, this essay shows that our understanding of seeming constants – whether they be the physical properties of water or the passions of love – can be modulated through attention to the specific histories of cognition and of embodiment.

  8. Melting method for radioactive solid wastes and device therefor

    Energy Technology Data Exchange (ETDEWEB)

    Komatsu, Masahiko; Abe, Takashi; Nakayama, Junpei; Kusamichi, Tatsuhiko; Sakamoto, Koichi

    1998-11-17

    Upon melting radioactive solid wastes mixed with radioactive metal wastes and non metal materials such as concrete by cold crucible high frequency induction heating, induction coils are wound around the outer circumference of a copper crucible having a water cooling structure to which radioactive solid wastes are charged. A heating sleeve formed by a material which generates heat by an induction heating function of graphite is disposed to the inside of the crucible at a height not in contact with molten metals in the crucible vertically movably. Radioactive solid wastes are melted collectively by the induction heat of the induction coils and thermal radiation and heat conduction of the heating sleeve heated by the induction heat. With such procedures, non metal materials such as concrete and radioactive metal wastes in a mixed state can be melt collectively continuously highly economically. (T.M.)

  9. Material properties influence on steam explosion efficiency. Prototypic versus simulant melts, eutectic versus non-eutectic melts

    International Nuclear Information System (INIS)

    Leskovar, M.; Mavko, B.

    2006-01-01

    A steam explosion may occur during a severe nuclear reactor accident if the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. Details of processes taking place prior and during the steam explosion have been experimentally studied for a number of years with adjunct efforts in modelling these processes to address the scaling of these experiments. Steam explosion experiments have shown that there are important differences of behaviour between simulant and prototypical melts, and that also at prototypical melts the fuel coolant interactions depend on the composition of the corium. In experiments with prototypic materials no spontaneous steam explosions occurred (except with an eutectic composition), whereas with simulant materials the steam explosions were triggered spontaneously. The energy conversion ratio of steam explosions with prototypic melts is at least one order of magnitude lower than the energy conversion ratio of steam explosions with simulant melts. Although the different behaviour of prototypic and simulant melts has been known for a number of years, there is no reliable explanation for these differences. Consequently it is not possible to reliably estimate whether corium would behave so non-explosive also in reactor conditions, where the mass of poured melt is nearly three orders of magnitude larger than in experimental conditions. An even more fascinating material effect was observed recently at corium experiments with eutectic and non-eutectic compositions. It turned out that eutectic corium always exploded spontaneously, whereas non-eutectic corium never exploded spontaneously. In the paper, a possible explanation of both material effects (prototypic/simulant melts, eutectic/non-eutectic corium) on the steam explosion is provided. A model for the calculation of the

  10. On high-pressure melting of tantalum

    Science.gov (United States)

    Luo, Sheng-Nian; Swift, Damian C.

    2007-01-01

    The issues related to high-pressure melting of Ta are discussed within the context of diamond-anvil cell (DAC) and shock wave experiments, theoretical calculations and common melting models. The discrepancies between the extrapolations of the DAC melting curve and the melting point inferred from shock wave experiments, cannot be reconciled either by superheating or solid-solid phase transition. The failure to reproduce low-pressure DAC melting curve by melting models such as dislocation-mediated melting and the Lindemann law, and molecular dynamics and quantum mechanics-based calculations, undermines their predictions at moderate and high pressures. Despite claims to the contrary, the melting curve of Ta (as well as Mo and W) remains inconclusive at high pressures.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  13. ASTEC application to in-vessel corium retention

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  14. On melting of boron phosphide under pressure

    OpenAIRE

    Solozhenko, Vladimir; Mukhanov, V. A.

    2015-01-01

    Melting of cubic boron phosphide, BP, has been studied at pressures to 9 GPa using synchrotron X-ray diffraction and electrical resistivity measurements. It has been found that above 2.6 GPa BP melts congruently, and the melting curve exhibits negative slope (–60 ± 7 K/GPa), which is indicative of a higher density of the melt as compared to the solid phase.

  15. Petrological Geodynamics of Mantle Melting I. AlphaMELTS + Multiphase Flow: Dynamic Equilibrium Melting, Method and Results

    Directory of Open Access Journals (Sweden)

    Massimiliano Tirone

    2017-10-01

    Full Text Available The complex process of melting in the Earth's interior is studied by combining a multiphase numerical flow model with the program AlphaMELTS which provides a petrological description based on thermodynamic principles. The objective is to address the fundamental question of the effect of the mantle and melt dynamics on the composition and abundance of the melt and the residual solid. The conceptual idea is based on a 1-D description of the melting process that develops along an ideal vertical column where local chemical equilibrium is assumed to apply at some level in space and time. By coupling together the transport model and the chemical thermodynamic model, the evolution of the melting process can be described in terms of melt distribution, temperature, pressure and solid and melt velocities but also variation of melt and residual solid composition and mineralogical abundance at any depth over time. In this first installment of a series of three contributions, a two-phase flow model (melt and solid assemblage is developed under the assumption of complete local equilibrium between melt and a peridotitic mantle (dynamic equilibrium melting, DEM. The solid mantle is also assumed to be completely dry. The present study addresses some but not all the potential factors affecting the melting process. The influence of permeability and viscosity of the solid matrix are considered in some detail. The essential features of the dynamic model and how it is interfaced with AlphaMELTS are clearly outlined. A detailed and explicit description of the numerical procedure should make this type of numerical models less obscure. The general observation that can be made from the outcome of several simulations carried out for this work is that the melt composition varies with depth, however the melt abundance not necessarily always increases moving upwards. When a quasi-steady state condition is achieved, that is when melt abundance does not varies significantly

  16. Design evolution and integration of the ITER in-vessel components

    International Nuclear Information System (INIS)

    Martin, A.; Calcagno, B.; Chappuis, Ph.; Daly, E.; Dellopoulos, G.; Furmanek, A.; Gicquel, S.; Heitzenroeder, P.; Jiming, Chen; Kalish, M.; Kim, D.-H.; Khomiakov, S.; Labusov, A.; Loarte, A.; Loughlin, M.; Merola, M.; Mitteau, R.; Polunovski, E.; Raffray, R.; Sadakov, S.

    2013-01-01

    Highlights: ► The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. ► A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. ► The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. ► The blanket manifold system has been redesigned to improve leak detection and localisation. ► The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. -- Abstract: The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues

  17. Filament stretching rheometry of polymer melts

    DEFF Research Database (Denmark)

    Hassager, Ole; Nielsen, Jens Kromann; Rasmussen, Henrik Koblitz

    2005-01-01

    The Filament Stretching Rheometry (FSR) method developed by Sridhar, McKinley and coworkers for polymer solutions has been extended to be used also for polymer melts. The design of a melt-FSR will be described and differences to conventional melt elongational rheometers will be pointed out. Results...

  18. Load histories from steam explosions during core melt accidents

    International Nuclear Information System (INIS)

    Jacobs, H.; Kolev, N.I.

    1992-01-01

    For the analysis of steam explosions a multicomponent multiphase thermohydraulic code is required which describes at least the motions of melt, water, and steam by separate velocity fields. One example of these very rare codes is the IVA3 code the development of which was brought to an interim close in 1991. As an example of a typical application of this code, precalculations of the FARO LWR Scoping Test 2 performed at Ispra are discussed. Unfortunately, the calculation results cannot be compared directly to the test results because of important differences between planned and achieved test parameters. Above all, only about one third of the planned melt mass actually entered the water. The test was performed in a closed vessel at an initial pressure of 50 bar. The water was saturated at this temperature and its level was at 1 m height. The simulation starts with the release of 50 kg of simulated corium from an intermediate catcher at about 3.2 m height. The calculation predicts a gradual pressure rise without fast transients worth mentioning from 50 to about 76 bar within roughly one second and stabilizes slightly below the maximum. Also described are the material distributions predicted during the process and the 'mixed' masses according to two different criteria. The former indicate that the melt jet penetrates the water without desintegrating while being surrounded by a thick vapor layer. Subsequently the melt collects at the level bottom and much of the liquid water is blown upwards by the steam being produced. The amounts of mass being 'mixed' with liquid water (and thus are thought to potentially participate in a steam explosion) remain below 10% for the known Theofanous criterion and below 30% for a more conservative criterion. It is however more important that the calculation demonstrates that further mixing could be the result of the onset of a steam explosion. This may strongly limit the usefulness of local mixing criteria. (orig./DG)

  19. Melt cooling by bottom flooding. The COMET core-catcher concept

    International Nuclear Information System (INIS)

    Foit, Jerzy Jan; Alsmeyer, Hans; Tromm, Walter; Buerger, Manfred; Journeau, Christophe

    2009-01-01

    The COMET concept has been developed to cool an ex-vessel corium melt in case of a hypothetical severe accident leading to vessel melt-through. After erosion of a sacrificial concrete layer the melt is passively flooded by bottom injection of coolant water. The open porosities and large surface that are generated during melt solidification form a porous permeable structure that is permanently filled with the evaporating water and thus allows an efficient short-term as well as long-term removal of the decay heat. The advantages of this concept are the fast cool-down and complete solidification of the melt within less than one hour typically. This stops further release of fission products from the corium. A drawback may be the fast release of steam during the quenching process. Several experimental series have been performed by FZK (Germany) to test and optimise the functionality of the different variants of the COMET concept. Thermite generated melts of iron and aluminium oxide were used. The large scale COMET-H test series with sustained inductive heating includes nine experiments performed with an array of water injection channels embedded in a sacrificial concrete layer. Variation of the water inlet pressure and melt height showed that melts up to 50 cm height can be safely cooled with an overpressure of the coolant water of 0.2 bar. The CometPC concept is based on cooling by flooding the melt from the bottom through layers of porous, water filled concrete. The third variant of the COMET design, CometPCA, uses a layer of porous, water filled concrete CometPCA from which flow channels protrude into the layer of sacrificial concrete. This modified concept combines the advantages of the original COMET concept with flow channels and the high resistance of a water-filled porous concrete layer against downward melt attack. Four large scale CometPCA experiments (FZK, Germany) have demonstrated an efficient cooling of melts up to 50 cm height using the recommended water

  20. Supercoil Formation During DNA Melting

    Science.gov (United States)

    Sayar, Mehmet; Avsaroglu, Baris; Kabakcioglu, Alkan

    2009-03-01

    Supercoil formation plays a key role in determining the structure-function relationship in DNA. Biological and technological processes, such as protein synthesis, polymerase chain reaction, and microarrays relys on separation of the two strands in DNA, which is coupled to the unwinding of the supercoiled structure. This problem has been studied theoretically via Peyrard-Bishop and Poland-Scheraga type models, which include a simple representation of the DNA structural properties. In recent years, computational models, which provide a more realtistic representaion of DNA molecule, have been used to study the melting behavior of short DNA chains. Here, we will present a new coarse-grained model of DNA which is capable of simulating sufficiently long DNA chains for studying the supercoil formation during melting, without sacrificing the local structural properties. Our coarse-grained model successfully reproduces the local geometry of the DNA molecule, such as the 3'-5' directionality, major-minor groove structure, and the helical pitch. We will present our initial results on the dynamics of supercoiling during DNA melting.

  1. Control Carbon to Prevent corium Stratification In-Vessel Retention

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

  2. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M.

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  3. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  4. Experiments and analyses on melt jet impingement during severe accidents

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Green, J.A.; Dinh, T.N.; Dong, W.

    1997-01-01

    Relocation of melt from the core region, during a nuclear reactor severe accident, presents the potential for erosion of the reactor pressure vessel (RPV) wall as a result of melt jet impingement. The extent of vessel erosion will depend upon a variety of parameters, including jet diameter, velocity, composition, superheat, angle of inclination, and the presence of an overlying water or melt pool. Experiments have been conducted at the Royal Institute of Technology Division of Nuclear Power Safety (RIT/NPS) which employ a variety of melt and pressure vessel simulant materials, such as water, salt-ice, Cerrobend alloy and molten salt. These experiments have revealed that the erosion depth of the vessel simulant in the jet stagnation zone can be adequately predicted by the Saito correlation, which is based on turbulent heat transfer, while initial erosion rates are seen to be in line with the laminar-stagnation-zone model. A transition between the laminar and turbulent regimes was realized in most cases and is attributed to the roughness of the surface in the eroded cavity formed

  5. EPRTM engineered features for core melt mitigation in severe accidents

    International Nuclear Information System (INIS)

    Fischer, Manfred; Henning, Andreas

    2009-01-01

    For the prevention of accident conditions, the EPR TM relies on the proven 3-level safety concepts inherited from its predecessors, the French 'N4' and the German 'Konvoi' NPP. In addition, a new, fourth 'beyond safety' level is implemented for the mitigation of postulated severe accidents (SA) with core melting. It is aimed at preserving the integrity of the containment barrier and at significantly reducing the frequency and magnitude of activity releases into the environment under such extreme conditions. Loss of containment integrity is prevented by dedicated design measures that address short- and long-term challenges, like: the melt-through of the reactor pressure vessel under high internal pressure, energetic hydrogen/steam explosions, containment overpressure failure, and basemat melt-through. The EPR TM SA systems and components that address these issues are: - the dedicated SA valves for the depressurization the primary circuit, - the provisions for H 2 recombination, atmospheric mixing, steam dilution, - the core melt stabilization system, - the dedicated SA containment heat removal system. The core melt stabilization system (CMSS) of the EPR TM is based on a two-stage ex-vessel approach. After its release from the RPV the core debris is first accumulated and conditioned in the (dry) reactor pit by the addition of sacrificial concrete. Then the created molten pool is spread into a lateral core catcher to establish favorable conditions for the later flooding, quenching and cooling with water passively drained from the Internal Refueling Water Storage Tank. Long-term heat removal from the containment is achieved by sprays that are supplied with water by the containment heat removal system. Complementing earlier publications focused on the principle function, basic design, and validation background of the EPR TM CMSS, this paper describes the state achieved after detailed design, as well as the technical solutions chosen for its main components, including

  6. Results of recent KROTOS FCI tests. Alumina vs. corium melts

    Energy Technology Data Exchange (ETDEWEB)

    Huhtiniemi, I.; Magallon, D.; Hohmann, H. [Commission of the European Communities, Ispra (Italy). Joint Research Center

    1998-01-01

    Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150 K - 750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1 - 0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed. (author)

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  8. Modeling of melt retention in EU-APR1400 ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Granovsky, V. S.; Sulatsky, A. A.; Khabensky, V. B.; Sulatskaya, M. B. [Alexandrov Research Inst. of Technology NITI, Sosnovy Bor (Russian Federation); Gusarov, V. V.; Almyashev, V. I.; Komlev, A. A. [Saint Petersburg State Technological Univ. SPbSTU, St.Petersburg (Russian Federation); Bechta, S. [KTH, Stockholm (Sweden); Kim, Y. S. [KHNP, 1312 Gil 70, Yuseongdaero, Yuseong-gu, Daejeon (Korea, Republic of); Park, R. J.; Kim, H. Y.; Song, J. H. [KAERI, 989 Gil 111, Daedeokdaero, Yuseong-gu, Daejeon (Korea, Republic of)

    2012-07-01

    A core catcher is adopted in the EU-APR1400 reactor design for management and mitigation of severe accidents with reactor core melting. The core catcher concept incorporates a number of engineering solutions used in the catcher designs of European EPR and Russian WER-1000 reactors, such as thin-layer corium spreading for better cooling, retention of the melt in a water-cooled steel vessel, and use of sacrificial material (SM) to control the melt properties. SM is one of the key elements of the catcher design and its performance is critical for melt retention efficiency. This SM consists of oxide components, but the core catcher also includes sacrificial steel which reacts with the metal melt of the molten corium to reduce its temperature. The paper describes the required properties of SM. The melt retention capability of the core catcher can be confirmed by modeling the heat fluxes to the catcher vessel to show that it will not fail. The fulfillment of this requirement is demonstrated on the example of LBLOCA severe accident. Thermal and physicochemical interactions between the oxide and metal melts, interactions of the melts with SM, sacrificial steel and vessel, core catcher external cooling by water and release of non-condensable gases are modeled. (authors)

  9. Industrial opportunities of controlled melt flow during glass melting, part 1: Melt flow evaluation

    Czech Academy of Sciences Publication Activity Database

    Dyrčíková, Petra; Hrbek, Lukáš; Němec, Lubomír

    2014-01-01

    Roč. 58, č. 2 (2014), s. 111-117 ISSN 0862-5468 R&D Projects: GA TA ČR TA01010844 Institutional support: RVO:67985891 Keywords : glass melting * controlled flow * space utilization Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 0.435, year: 2014 http://www.ceramics-silikaty.cz/2014/pdf/2014_02_111.pdf

  10. Synergistic effect of fly ash in in-vessel composting of biomass and kitchen waste.

    Science.gov (United States)

    Manyapu, Vivek; Mandpe, Ashootosh; Kumar, Sunil

    2018-03-01

    The present study aims to utilize coal fly ash for its property to adsorb heavy metals and thus reducing the bioavailability of the metals for plant uptake. Fly ash was incorporated into the in-vessel composting system along with organic waste. The in-vessel composting experiments were conducted in ten plastic vessels of 15 L capacity comprising varying proportions of biomass waste, kitchen waste and fly ash. In this study, maximum degradation of organic matter was observed in Vessel 3 having k value of 0.550 d -1 . In vessel 10, 20% fly ash with a combination of 50% biomass waste and 30% kitchen waste along with the addition of 5% jaggery as an additive produced the best outcome with least organic matter (%C) loss and lowest value of rate constant (k). Copyright © 2017 Elsevier Ltd. All rights reserved.

  11. Eutectic melting temperature of the lowermost Earth's mantle

    Science.gov (United States)

    Andrault, D.; Lo Nigro, G.; Bolfan-Casanova, N.; Bouhifd, M.; Garbarino, G.; Mezouar, M.

    2009-12-01

    , and changes in the relation between sample-temperature and laser-power. In this work, we show that temperatures higher than 4000 K are necessary for melting mean mantle at the 135 GPa pressure found at the core mantle boundary (CMB). Such temperature is much higher than that from estimated actual geotherms. Therefore, melting at the CMB can only occur if (i) pyrolitic mantle resides for a very long time in contact with the outer core, (ii) the mantle composition is severely affected by additional elements depressing the solidus such as water or (iii) the temperature gradient in the D" region is amazingly steep. Other implications for the temperature state and the lower mantle properties will be presented. References (1) Ito et al., Phys. Earth Planet. Int., 143-144, 397-406, 2004 (2) Ohtani et al., Phys. Earth Planet. Int., 100, 97-114, 1997 (3) Zerr et al., Science, 281, 243-246, 1998 (4) Holland and Ahrens, Science, 275, 1623-1625, 1997 (5) Schultz et al., High Press. Res., 25, 1, 71-83, 2005.

  12. On heat transfer characteristics of real and simulant melt pool experiments

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Nourgaliev R.R.; Sehgal, B.R. [Royal Institute of Technology, Stockholm (Sweden)

    1995-09-01

    The paper presents results of analytical studies of natural convection heat transfer in scaled and/or simulant melt pool experiments related to the PWR in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing physical processes under prototypical situations is examined. Experiments and experimental approaches are analysed by focusing on their ability to represent prototypical situations. Calculations are carried out in order to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. Rayleigh numbers in the present analysis are limited to 10{sup 12}, where uncertainties in turbulence modeling are not overriding other uncertainties. The effects of fluid Prandtl number on heat transfer to the lowermost part of cooled pool walls are examined for square and semicircular cavities. Calculations are performed also to explore limitations of using side-wall heating and direct electrical heating in reproducing the physical picture of interest. Needs for further experimental and analytical efforts are discussed as well.

  13. How much can Greenland melt? An upper bound on mass loss from the Greenland Ice Sheet through surface melting

    Science.gov (United States)

    Liu, X.; Bassis, J. N.

    2015-12-01

    With observations showing accelerated mass loss from the Greenland Ice Sheet due to surface melt, the Greenland Ice Sheet is becoming one of the most significant contributors to sea level rise. The contribution of the Greenland Ice Sheet o sea level rise is likely to accelerate in the coming decade and centuries as atmospheric temperatures continue to rise, potentially triggering ever larger surface melt rates. However, at present considerable uncertainty remains in projecting the contribution to sea level of the Greenland Ice Sheet both due to uncertainty in atmospheric forcing and the ice sheet response to climate forcing. Here we seek an upper bound on the contribution of surface melt from the Greenland to sea level rise in the coming century using a surface energy balance model coupled to an englacial model. We use IPCC Representative Concentration Pathways (RCP8.5, RCP6, RCP4.5, RCP2.6) climate scenarios from an ensemble of global climate models in our simulations to project the maximum rate of ice volume loss and related sea-level rise associated with surface melting. To estimate the upper bound, we assume the Greenland Ice Sheet is perpetually covered in thick clouds, which maximize longwave radiation to the ice sheet. We further assume that deposition of black carbon darkens the ice substantially turning it nearly black, substantially reducing its albedo. Although assuming that all melt water not stored in the snow/firn is instantaneously transported off the ice sheet increases mass loss in the short term, refreezing of retained water warms the ice and may lead to more melt in the long term. Hence we examine both assumptions and use the scenario that leads to the most surface melt by 2100. Preliminary models results suggest that under the most aggressive climate forcing, surface melt from the Greenland Ice Sheet contributes ~1 m to sea level by the year 2100. This is a significant contribution and ignores dynamic effects. We also examined a lower bound

  14. Dynamic simulation of a planar flexible boom for tokamak in-vessel operations

    International Nuclear Information System (INIS)

    Ambrosino, G.; Celentano, G.; Garofalo, F.; Maisonnier, D.

    1991-01-01

    In this paper we present a dynamic model for the analysis of the vibrations of a planar articulated flexible boom to be used for tokamak in-vessel maintenance operations. The peculiarity of the mechanical structure of the boom enables us to consider separately the oscillations in the horizontal and vertical planes so that two separate models can be constructed for describing these phenomena. The results of simulations based on booms like that proposed for NET in-vessel operations are presented. (orig.)

  15. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  16. Simulation experiment on the flooding behaviour of core melts: KATS-9

    International Nuclear Information System (INIS)

    Fieg, G.; Massier, H.; Schuetz, W.; Stegmaier, U.; Stern, G.

    2000-11-01

    For future Light Water Reactors special devices (core catchers) are being developed to prevent containment failure by basement erosion after reactor pressure vessel meltthrough during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher devices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent water cooling. A KATS-experiment has been performed to investigate the flooding behaviour of high temperature melts using alumina-iron thermite melts as a simulant. The oxidic thermite melt is conditioned by adding other oxides to simulate a realistic corium melt as close as possible in terms of liquidus and solidus temperatures. Before flooding with water, spreading of the separate oxidic and metallic melts has been done in one-dimensional channels with a silicate concrete as the substrate. The flooding rate was, in relation to the melt surface, identical to the flooding rate in EPR. (orig.) [de

  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. The impact of melt ponds on summertime microwave brightness temperatures and sea-ice concentrations

    DEFF Research Database (Denmark)

    Kern, Stefan; Rösel, Anja; Pedersen, Leif Toudal

    2016-01-01

    % sea-ice concentration. None of the algorithms investigated performs best based on our investigation of data from summer 2009. We suggest that those algorithms which are more sensitive to melt ponds could be optimized more easily because the influence of unknown snow and sea-ice surface property...... of eight sea-ice concentration retrieval algorithms to melt ponds by comparing sea-ice concentration with the melt-pond fraction. We derive gridded daily sea-ice concentrations from microwave brightness temperatures of summer 2009. We derive the daily fraction of melt ponds, open water between ice floes......, and the ice-surface fraction from contemporary Moderate Resolution Spectroradiometer (MODIS) reflectance data. We only use grid cells where the MODIS sea ice concentration, which is the melt-pond fraction plus the ice-surface fraction, exceeds 90 %. For one group of algorithms, e.g., Bristol and Comiso...

  19. Melt refining of uranium contaminated copper, nickel, and mild steel

    International Nuclear Information System (INIS)

    Ren Xinwen; Liu Wencang; Zhang Yuan

    1993-01-01

    This paper presents the experiment results on melt refining of uranium contaminated metallic discards such as copper, nickel, and mild steel. Based on recommended processes, uranium contents in ingots shall decrease below 1 ppm; metal recovery is higher than 96%; and slag production is below 5% in weight of the metal to be refined. The uranium in the slag is homogeneously distributed. The slag seems to be hard ceramics, insoluble in water, and can be directly disposed of after proper packaging

  20. Scattering of light at the growing solid-melt interface

    International Nuclear Information System (INIS)

    Gontijo, I.

    1987-12-01

    The scattering of light at the growing solid-melt interface of biphenyl and naphthalene was studied using the Photon Correlation Spectroscopy technique. The origin of this light scattering remained without a satisfactory explanation since its discovery at the ice-water interface in 1978. Recently, a model based on the segregation of gaseous impurities at the interface and subsequent precipitation of microbubbles was proposed to explain this phenomenon. We report here the first experimental results that confirm the microbbubles hypothesis. (author)

  1. An experimental study of steam explosions involving CORIUM melts

    International Nuclear Information System (INIS)

    Millington, R.A.

    1984-05-01

    An experimental programme to investigate molten fuel coolant interactions involving 0.5 kg thermite-generated CORIUM melts and water has been carried out. System pressures and initial coolant subcoolings were chosen to enhance the probability of steam explosions. Yields and efficiencies of the interactions were found to be very close to those obtained from similar experiments using molten UO 2 generated from a Uranium/Molybdenum Trioxide thermite. (author)

  2. Double melting in polytetrafluoroethylene γ-irradiated above its melting point

    International Nuclear Information System (INIS)

    Serov, S.A.; Khatipov, S.A.; Sadovskaya, N.V.; Tereshenkov, A.V.; Chukov, N.A.

    2012-01-01

    Highlights: ► PTFE irradiation leads to formation of double melting peaks in DSC curves. ► This is connected to dual crystalline morphology typical for PTFE. ► Two crystalline types exist in the PTFE irradiated in the melt. - Abstract: PTFE irradiation above its melting point leads to formation of double melting and crystallization peaks in DSC curves. Splitting of melting peaks is connected to dual crystalline morphology typical for PTFE irradiated in the melt. According to electron microscopy, two crystalline types with different size and packing density exist in the irradiated PTFE.

  3. Tin in granitic melts: The role of melting temperature and protolith composition

    Science.gov (United States)

    Wolf, Mathias; Romer, Rolf L.; Franz, Leander; López-Moro, Francisco Javier

    2018-06-01

    Granite bound tin mineralization typically is seen as the result of extreme magmatic fractionation and late exsolution of magmatic fluids. Mineralization, however, also could be obtained at considerably less fractionation if initial melts already had enhanced Sn contents. We present chemical data and results from phase diagram modeling that illustrate the dominant roles of protolith composition, melting conditions, and melt extraction/evolution for the distribution of Sn between melt and restite and, thus, the Sn content of melts. We compare the element partitioning between leucosome and restite of low-temperature and high-temperature migmatites. During low-temperature melting, trace elements partition preferentially into the restite with the possible exception of Sr, Cd, Bi, and Pb, that may be enriched in the melt. In high-temperature melts, Ga, Y, Cd, Sn, REE, Pb, Bi, and U partition preferentially into the melt whereas Sc, V, Cr, Co, Ni, Mo, and Ba stay in the restite. This contrasting behavior is attributed to the stability of trace element sequestering minerals during melt generation. In particular muscovite, biotite, titanite, and rutile act as host phases for Sn and, therefore prevent Sn enrichment in the melt as long as they are stable phases in the restite. As protolith composition controls both the mineral assemblage and modal contents of the various minerals, protolith composition eventually also controls the fertility of a rock during anatexis, restite mineralogy, and partitioning behavior of trace metals. If a particular trace element is sequestered in a phase that is stable during partial melting, the resulting melt is depleted in this element whereas the restite becomes enriched. Melt generation at high temperature may release Sn when Sn-hosts become unstable. If melt has not been lost before the breakdown of Sn-hosts, Sn contents in the melt will increase but never will be high. In contrast, if melt has been lost before the decomposition of Sn

  4. Chemical decontamination and melt densification

    International Nuclear Information System (INIS)

    Dillon, R.L.; Griggs, B.; Kemper, R.S.; Nelson, R.G.

    1976-01-01

    Preliminary studies on the chemical decontamination and densification of Zircaloy, stainless steel, and Inconel undissolved residues remaining after dissolution of the UO 2 --PuO 2 spent fuel material from sheared fuel bundles are reported. The studies were made on cold or very small samples to demonstrate the feasibility of the processes developed before proceeding to hot cell demonstrations with kg level of the sources. A promising aqueous decontamination method for Zr alloy cladding was developed in which oxidized surfaces are conditioned with HF prior to leaching with ammonium oxalate, ammonium citrate, ammonium fluoride, and hydrogen peroxide. Feasibility of molten salt decontamination of oxidized Zircaloy was demonstrated. A low melting alloy of Zircaloy, stainless steel, and Inconel was obtained in induction heated graphite crucibles. Segregated Zircaloy cladding sections were directly melted by the inductoslag process to yield a metal ingot suitable for storage. Both Zircaloy and Zircaloy--stainless steel--Inconel alloys proved to be highly satisfactory getters and sinks for recovered tritium

  5. Monitoring of polymer melt processing

    International Nuclear Information System (INIS)

    Alig, Ingo; Steinhoff, Bernd; Lellinger, Dirk

    2010-01-01

    The paper reviews the state-of-the-art of in-line and on-line monitoring during polymer melt processing by compounding, extrusion and injection moulding. Different spectroscopic and scattering techniques as well as conductivity and viscosity measurements are reviewed and compared concerning their potential for different process applications. In addition to information on chemical composition and state of the process, the in situ detection of morphology, which is of specific interest for multiphase polymer systems such as polymer composites and polymer blends, is described in detail. For these systems, the product properties strongly depend on the phase or filler morphology created during processing. Examples for optical (UV/vis, NIR) and ultrasonic attenuation spectra recorded during extrusion are given, which were found to be sensitive to the chemical composition as well as to size and degree of dispersion of micro or nanofillers in the polymer matrix. By small-angle light scattering experiments, process-induced structures were detected in blends of incompatible polymers during compounding. Using conductivity measurements during extrusion, the influence of processing conditions on the electrical conductivity of polymer melts with conductive fillers (carbon black or carbon nanotubes) was monitored. (topical review)

  6. Feasibility of re-melting NORM-contaminated scrap metal

    Energy Technology Data Exchange (ETDEWEB)

    Winters, S. J.; Smith, K. P.

    1999-10-26

    Naturally occurring radioactive materials (NORM) sometimes accumulate inside pieces of equipment associated with oil and gas production and processing activities. Typically, the NORM accumulates when radium that is present in solution in produced water precipitates out in scale and sludge deposits. Scrap equipment containing residual quantities of these NORM-bearing scales and sludges can present a waste management problem if the radium concentrations exceed regulatory limits or activate the alarms on radiation screening devices installed at most scrap metal recycling facilities. Although NORM-contaminated scrap metal currently is not disposed of by re-melting, this form of recycling could present a viable disposition option for this waste stream. Studies indicate that re-melting NORM-contaminated scrap metal is a viable recycling option from a risk-based perspective. However, a myriad of economic, regulatory, and policy issues have caused the recyclers to turn away virtually all radioactive scrap metal. Until these issues can be resolved, re-melting of the petroleum industry's NORM-impacted scrap metal is unlikely to be a widespread practice. This paper summarizes the issues associated with re-melting radioactive scrap so that the petroleum industry and its regulators will understand the obstacles. This paper was prepared as part of a report being prepared by the Interstate Oil and Gas Compact Commission's NORM Subcommittee.

  7. Studies of Behavior Melting Temperature Characteristics for Multi Thermocouple In-Core Instrument Assembly

    International Nuclear Information System (INIS)

    Shin, Donghyup; Chae, Myoungeun; Kim, Sungjin; Lee, Kyulim

    2015-01-01

    Bottom-up type in-core instruments (ICIs) are used for the pressurized water reactors of OPR-1000, APR- 1400 in order to measure neutron flux and temperature in the reactor. It is a well-known technique and a proven design using years in the nuclear field. ICI consists of one pair of K-type thermocouple, five self-powered neutron detectors (SPNDs) and one back ground detector. K-type thermocouple's purpose is to measure the core exit temperature (CET) in the reactor. The CET is a very important factor for operating nuclear power plants and it is 327 .deg. C when generally operating the reactor in the nuclear power plant(NPP) in case of OPR- 1000. If the CET will exceed 650 .deg. C, Operators in the main control room should be considered to be an accident situation in accordance with a severe accident management guidance(SAMG). The Multi Thermocouple ICI is a new designed ICI assuming severe accident conditions. It consists of four more thermocouples than the existing design, so it has five Ktype thermocouples besides the thermocouple measuring CET is located in the same elevation as the ICI. Each thermocouple is able to be located in the desired location as required. The Multi Thermocouple ICI helps to measure the temperature distribution of the entire reactor. In addition, it will measure certain point of melted core because of the in-vessel debris of nuclear fuel when an accident occurs more seriously. In this paper, to simulate a circumstance such as a nuclear reactor severe accident was examined. In this study, the K-type thermocouples of Multi Thermocouple ICI was confirmed experimentally to be able to measure up to 1370 .deg. C before the thermocouples have been melted. And after the thermocouples were melted by debris, it was able to be monitored that the signal of EMF directed the infinite value of voltage. Therefore through the results of the test, it can be assumed that if any EMF data among the Multi Thermocouple ICI will direct the infinite value

  8. Studies of Behavior Melting Temperature Characteristics for Multi Thermocouple In-Core Instrument Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Donghyup; Chae, Myoungeun; Kim, Sungjin; Lee, Kyulim [Woojin inc, Hwasung (Korea, Republic of)

    2015-05-15

    Bottom-up type in-core instruments (ICIs) are used for the pressurized water reactors of OPR-1000, APR- 1400 in order to measure neutron flux and temperature in the reactor. It is a well-known technique and a proven design using years in the nuclear field. ICI consists of one pair of K-type thermocouple, five self-powered neutron detectors (SPNDs) and one back ground detector. K-type thermocouple's purpose is to measure the core exit temperature (CET) in the reactor. The CET is a very important factor for operating nuclear power plants and it is 327 .deg. C when generally operating the reactor in the nuclear power plant(NPP) in case of OPR- 1000. If the CET will exceed 650 .deg. C, Operators in the main control room should be considered to be an accident situation in accordance with a severe accident management guidance(SAMG). The Multi Thermocouple ICI is a new designed ICI assuming severe accident conditions. It consists of four more thermocouples than the existing design, so it has five Ktype thermocouples besides the thermocouple measuring CET is located in the same elevation as the ICI. Each thermocouple is able to be located in the desired location as required. The Multi Thermocouple ICI helps to measure the temperature distribution of the entire reactor. In addition, it will measure certain point of melted core because of the in-vessel debris of nuclear fuel when an accident occurs more seriously. In this paper, to simulate a circumstance such as a nuclear reactor severe accident was examined. In this study, the K-type thermocouples of Multi Thermocouple ICI was confirmed experimentally to be able to measure up to 1370 .deg. C before the thermocouples have been melted. And after the thermocouples were melted by debris, it was able to be monitored that the signal of EMF directed the infinite value of voltage. Therefore through the results of the test, it can be assumed that if any EMF data among the Multi Thermocouple ICI will direct the infinite value

  9. Criteria for the spreading of oxide melts: Test series miniKATS-1 to -5

    International Nuclear Information System (INIS)

    Eppinger, B.; Fieg, G.; Massier, H.; Schuetz, W.; Stegmaier, U.; Stern, G.

    2001-09-01

    In a long series of larger spreading tests with high temperature oxide melts (KATS tests) many parameters have been varied which are influencing the spreading behaviour (viscosity, pouring rate, substratum of spreading surface, presence of water). In spite of an extensive instrumentation using different thermocouples, an infrared camera and several video cameras, only in very few cases the behaviour of the melt front at the very moment of immobilization could be detected in detail. Therefore in the additional miniKATS series five small scale (5 kg) spreading tests with oxide melts have been conducted to investigate the mechanical properties of the spreading front in the moment of immobilization. It turned out that in all cases the bulk of the melt at this moment was still liquid at a temperature close to the initial one. Depending on the initial melt properties two distinct phenomena have been observed which control the immobilization of the melt: the first phenomena is the crust formation at the surface, the crusts at the bottom combined with the crust at the surface of the melt. In the other case the whole melt front was still above the liquid temperature at the moment of immobilization. Here the surface tension was controlling the spreading, it was in balance with the driving gravitational force. In none of the test bulk freezing has been detected. (orig.)

  10. Variability of Basal Melt Beneath the Pine Island Glacier Ice Shelf, West Antarctica

    Science.gov (United States)

    Bindschadler, Robert; Vaughan, David G.; Vornberger, Patricia

    2011-01-01

    Observations from satellite and airborne platforms are combined with model calculations to infer the nature and efficiency of basal melting of the Pine Island Glacier ice shelf, West Antarctica, by ocean waters. Satellite imagery shows surface features that suggest ice-shelf-wide changes to the ocean s influence on the ice shelf as the grounding line retreated. Longitudinal profiles of ice surface and bottom elevations are analyzed to reveal a spatially dependent pattern of basal melt with an annual melt flux of 40.5 Gt/a. One profile captures a persistent set of surface waves that correlates with quasi-annual variations of atmospheric forcing of Amundsen Sea circulation patterns, establishing a direct connection between atmospheric variability and sub-ice-shelf melting. Ice surface troughs are hydrostatically compensated by ice-bottom voids up to 150m deep. Voids form dynamically at the grounding line, triggered by enhanced melting when warmer-than-average water arrives. Subsequent enlargement of the voids is thermally inefficient (4% or less) compared with an overall melting efficiency beneath the ice shelf of 22%. Residual warm water is believed to cause three persistent polynyas at the ice-shelf front seen in Landsat imagery. Landsat thermal imagery confirms the occurrence of warm water at the same locations.

  11. Features of melting of indium monohalides

    Energy Technology Data Exchange (ETDEWEB)

    Dmitriev, V S; Smirniv, V A [AN SSSR, Chernogolovka. Inst. Fiziki Tverdogo Tela

    1980-12-01

    The character of InCl, InBr and InI melting is investigated by the methods of DTA, calorimetry, conductometry and chemical analysis. Partial decomposition of monohalogenides during melting according to the reactions of disproportionation is shown. The presence of disproportionation products (In/sup 0/ and In/sup 3 +/) is manifested in the properties of solid monohalogenides, prepared by the crystallization from melt, in their photosensitivity and electroconductivity.

  12. Multiscale Models of Melting Arctic Sea Ice

    Science.gov (United States)

    2014-09-30

    Sea ice reflectance or albedo , a key parameter in climate modeling, is primarily determined by melt pond and ice floe configurations. Ice - albedo ...determine their albedo - a key parameter in climate modeling. Here we explore the possibility of a conceptual sea ice climate model passing through a...bifurcation points. Ising model for melt ponds on Arctic sea ice Y. Ma, I. Sudakov, and K. M. Golden Abstract: The albedo of melting

  13. Reaction- and melting behaviour of LWR-core components UO2, Zircaloy and steel during the meltdown period

    International Nuclear Information System (INIS)

    Hofmann, P.

    1976-07-01

    The reaction behaviour of the UO 2 , Zircaloy-4 and austenitic steel core components was investigated as a function of temperature (till melting temperatures) under inert and oxidizing conditions. Component concentrations varied between that of Corium-A (65 wt.% UO 2 , 18% Zry, 17% steel) and that of Corium-E (35 wt.% UO 2 , 10% Zry, 55% steel). In addition, Zircaloy and stainless steel were used with different degrees of oxidation. The paper describes systematically the phases that arise during heating and melting. The integral composition of the melts and the qualitative as well as quantitative analysis of the phases present in solidified corium are given. In some cases melting points have been determined. The reaction and melting behaviour of the corium specimens strongly depends on the concentration and on the degree of oxidation of the core components. First liquid phases are formed at the Zry-steel interface at about 1,350 0 C. The maximum temperatures of about 2,500 0 C for the complete melting of the corium-specimens are well below the UO 2 melting point. Depending on the steel content and/or degree of oxidation of Zry and steel, a homogeneous metallic or oxide melt or two immiscible melts - one oxide and the other metallic - are obtained. During the melting experiments performed under inert gas conditions the chemical composition of the molten specimens generally change by evaporation losses of single elements, especially of uranium, zirconium and oxygen. The total weight losses go up to 30%; under oxidizing conditions they are substantially smaller due to the occurrence of different phases. In air or water vapor, the occurrence of the phases and the melting behaviour of the core components are strongly influenced by the oxidation rate and the oxygen supply to the surface of the melt. In the case of the hypothetical core melting accident, a heterogeneous melt (oxide and metallic) is probable after the meltdown period. (orig./RW) [de

  14. Calculation of melting points of oxides

    International Nuclear Information System (INIS)

    Bobkova, O.S.; Voskobojnikov, V.G.; Kozin, A.I.

    1975-01-01

    The correlation between the melting point and thermodynamic parameters characterizing the strength of oxides and compounds is given. Such thermodynamic paramters include the energy and antropy of atomization

  15. A 400-year ice core melt layer record of summertime warming in the Alaska Range

    Science.gov (United States)

    Winski, D.; Osterberg, E. C.; Kreutz, K. J.; Wake, C. P.; Ferris, D. G.; Campbell, S. W.; Baum, M.; Raudzens Bailey, A.; Birkel, S. D.; Introne, D.; Handley, M.

    2017-12-01

    Warming in high-elevation regions has socially relevant impacts on glacier mass balance, water resources, and sensitive alpine ecosystems, yet very few high-elevation temperature records exist from the middle or high latitudes. While many terrestrial paleoclimate records provide critical temperature records from low elevations over recent centuries, melt layers preserved in alpine glaciers present an opportunity to develop calibrated, annually-resolved temperature records from high elevations. We present a 400-year temperature record based on the melt-layer stratigraphy in two ice cores collected from Mt. Hunter in the Central Alaska Range. The ice core record shows a 60-fold increase in melt frequency and water equivalent melt thickness between the pre-industrial period (before 1850) and present day. We calibrate the melt record to summer temperatures based on local and regional weather station analyses, and find that the increase in melt production represents a summer warming of at least 2° C, exceeding rates of temperature increase at most low elevation sites in Alaska. The Mt. Hunter melt layer record is significantly (p<0.05) correlated with surface temperatures in the central tropical Pacific through a Rossby-wave like pattern that induces high temperatures over Alaska. Our results show that rapid alpine warming has taken place in the Alaska Range for at least a century, and that conditions in the tropical oceans contribute to this warming.

  16. Crack growth rates in vessel head penetration materials

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Blazquez, F.

    1994-01-01

    The cracks detected in reactor vessel head penetrations in certain European plants have been attributed to Primary Water Stress Corrosion Cracking (PWSCC). The penetrations in question are made from Inconel 600. The susceptibility of this alloy to PWSCC has been widely studied in relation to use of this material for steam generator tubes. When the first reactor vessel head penetration cracks were detected, most of the available data on crack propagation rates were from test specimens made from steam generator tubes and tested under conditions that questioned the validity of these data for assessment of the evolution of cracks in penetrations. For this reason, the scope of the Spanish Research Project on the Inspection and Repair of PWR reactor vessel head penetrations included the acquisition of data on crack propagation rates in Inconel 600, representative of the materials used for vessel head penetrations. (authors). 1 fig., 2 tabs., 6 refs

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  18. Comparative Study on Two Melting Simulation Methods: Melting Curve of Gold

    International Nuclear Information System (INIS)

    Liu Zhong-Li; Li Rui; Sun Jun-Sheng; Zhang Xiu-Lu; Cai Ling-Cang

    2016-01-01

    Melting simulation methods are of crucial importance to determining melting temperature of materials efficiently. A high-efficiency melting simulation method saves much simulation time and computational resources. To compare the efficiency of our newly developed shock melting (SM) method with that of the well-established two-phase (TP) method, we calculate the high-pressure melting curve of Au using the two methods based on the optimally selected interatomic potentials. Although we only use 640 atoms to determine the melting temperature of Au in the SM method, the resulting melting curve accords very well with the results from the TP method using much more atoms. Thus, this shows that a much smaller system size in SM method can still achieve a fully converged melting curve compared with the TP method, implying the robustness and efficiency of the SM method. (paper)

  19. Commissioning result of the KSTAR in-vessel cryo-pump

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. B.; Lee, H. J.; Park, Y.M. [National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2013-12-15

    KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to LN2 one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

  20. Development of Ultrasonic Visual Inspection Program for In-Vessel Structures of SFR

    International Nuclear Information System (INIS)

    Joo, Y. S.; Park, C. G.; Lee, J. H.

    2009-02-01

    As the liquid sodium of a sodium-cooled fast reactor (SFR) is opaque to light, a conventional visual inspection is unavailable for the evaluation of the in-vessel structures under a sodium level. ASME Section XI Division 3 provides rules and guidelines for an in-service inspection (ISI) and testing of the components of SFR. For the ISI of in-vessel structures, the ASME code specifies visual examinations. An ultrasonic wave should be applied for an under-sodium visual inspection of the in-vessel structures. The plate-type waveguide sensor has been developed and the feasibility of the waveguide sensor technique has been successfully demonstrated for an ultrasonic visual inspection of the in-vessel structures of SFR. In this study, the C-scan image mapping program (Under-Sodium MultiView) is developed to apply this waveguide sensor technology to an under-sodium visual inspection of in-vessel structures in SFR by using a LabVIEW graphical programming language. The Under-Sodium MultiVIEW program has the functions of a double rotating scanner motion control, a high power pulser receiver control, a image mapping and a signal processing. The performance of Under-Sodium MultiVIEW program was verified by a C-scanning test

  1. Recycling melting process of the zirconium alloy chips

    International Nuclear Information System (INIS)

    Reis, Luis A.M. dos; Mucsi, Cristiano S.; Tavares, Luiz A.P.; Alencar, Maicon C.; Gomes, Maurilio P.; Barbosa, Luzinete P.; Rossi, Jesualdo L.

    2017-01-01

    Pressurized water reactors (PWR) commonly use 235 U enriched uranium dioxide pellets as a nuclear fuel, these are assembled and stacked in zirconium alloy tubes and end caps (M5, Zirlo, Zircaloy). During the machining of these components large amounts of chips are generated which are contaminated with cutting fluid. Its storage presents safety and environmental risks due to its pyrophoric and reactive nature. Recycling industry shown interest in its recycling due to its strategic importance. This paper presents a study on the recycling process and the results aiming the efficiency in the cleaning process; the quality control; the obtaining of the pressed electrodes and finally the melting in a Vacuum Arc Remelting furnace (VAR). The recycling process begins with magnetic separation of possible ferrous alloys chips contaminant, the washing of the cutting fluid that is soluble in water, washing with an industrial degreaser, followed by a rinse with continuous flow of water under high pressure and drying with hot air. The first evaluation of the process was done by an Energy Dispersive X-rays Fluorescence Spectrometry (EDXRFS) showed the presence of 10 wt. % to 17 wt. % of impurities due the mixing with stainless steel machining chips. The chips were then pressed in a custom-made matrix of square section (40 x 40 mm - 500 mm in length), resulting in electrodes with 20% of apparent density of the original alloy. The electrode was then melted in a laboratory scale VAR furnace at the CCTM-IPEN, producing a massive ingot with 0.8 kg. It was observed that the samples obtained from Indústrias Nucleares do Brasil (INB) are supposed to be secondary scrap and it is suggested careful separation in the generation of this material. The melting of the chips is possible and feasible in a VAR furnace which reduces the storage volume by up to 40 times of this material, however, it is necessary to correct the composition of the alloy for the melting of these ingots. (author)

  2. Percolation blockage: A process that enables melt pond formation on first year Arctic sea ice

    Science.gov (United States)

    Polashenski, Chris; Golden, Kenneth M.; Perovich, Donald K.; Skyllingstad, Eric; Arnsten, Alexandra; Stwertka, Carolyn; Wright, Nicholas

    2017-01-01

    Melt pond formation atop Arctic sea ice is a primary control of shortwave energy balance in the Arctic Ocean. During late spring and summer, the ponds determine sea ice albedo and how much solar radiation is transmitted into the upper ocean through the sea ice. The initial formation of ponds requires that melt water be retained above sea level on the ice surface. Both theory and observations, however, show that first year sea ice is so highly porous prior to the formation of melt ponds that multiday retention of water above hydraulic equilibrium should not be possible. Here we present results of percolation experiments that identify and directly demonstrate a mechanism allowing melt pond formation. The infiltration of fresh water into the pore structure of sea ice is responsible for blocking percolation pathways with ice, sealing the ice against water percolation, and allowing water to pool above sea level. We demonstrate that this mechanism is dependent on fresh water availability, known to be predominantly from snowmelt, and ice temperature at melt onset. We argue that the blockage process has the potential to exert significant control over interannual variability in ice albedo. Finally, we suggest that incorporating the mechanism into models would enhance their physical realism. Full treatment would be complex. We provide a simple temperature threshold-based scheme that may be used to incorporate percolation blockage behavior into existing model frameworks.

  3. Molecular dynamics simulations of melting behavior of alkane as phase change materials slurry

    International Nuclear Information System (INIS)

    Rao Zhonghao; Wang Shuangfeng; Wu Maochun; Zhang Yanlai; Li Fuhuo

    2012-01-01

    Highlights: ► The melting behavior of phase change materials slurry was investigated by molecular dynamics simulation method. ► Four different PCM slurry systems including pure water and water/n-nonadecane composite were constructed. ► Amorphous structure and periodic boundary conditions were used in the molecular dynamics simulations. ► The simulated melting temperatures are very close to the published experimental values. - Abstract: The alkane based phase change materials slurry, with high latent heat storage capacity, is effective to enhance the heat transfer rate of traditional fluid. In this paper, the melting behavior of composite phase change materials slurry which consists of n-nonadecane and water was investigated by using molecular dynamics simulation. Four different systems including pure water and water/n-nonadecane composite were constructed with amorphous structure and periodic boundary conditions. The results showed that the simulated density and melting temperature were very close to the published experimental values. Mixing the n-nonadecane into water decreased the mobility but increased the energy storage capacity of composite systems. To describe the melting behavior of alkane based phase change materials slurry on molecular or atomic scale, molecular dynamics simulation is an effective method.

  4. Modelling of the controlled melt flow in a glass melting space – Its melting performance and heat losses

    Czech Academy of Sciences Publication Activity Database

    Jebavá, Marcela; Dyrčíková, Petra; Němec, Lubomír

    2015-01-01

    Roč. 430, DEC 15 (2015), s. 52-63 ISSN 0022-3093 Institutional support: RVO:67985891 Keywords : glass melt flow * mathematical modelling * energy distribution * space utilizatios * melting performance Subject RIV: JH - Ceramics, Fire-Resistant Materials and Glass Impact factor: 1.825, year: 2015

  5. Melting and Sintering of Ashes

    DEFF Research Database (Denmark)

    Hansen, Lone Aslaug

    1997-01-01

    -1300°C, and a trend of higher fusion temperatures with increasing contents of Al-silicates and quartz was found.c) Fly ashes, bottom ashes and deposits from coal/straw co-firing were all found to consist mainly of metal-alumina and alumina-silicates. These ashes all melt in the temperature range 1000......The thesis contains an experimental study of the fusion and sintering of ashes collected during straw and coal/straw co-firing.A laboratory technique for quantitative determination of ash fusion has been developed based on Simultaneous Thermal Analysis (STA). By means of this method the fraction......, the biggest deviations being found for salt rich (i.e. straw derived) ashes.A simple model assuming proportionality between fly ash fusion and deposit formation was found to be capable of ranking deposition rates for the different straw derived fly ashes, whereas for the fly ashes from coal/straw co-firing...

  6. 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

  7. High-pressure melting curve of KCl: Evidence against lattice-instability theories of melting

    International Nuclear Information System (INIS)

    Ross, M.; Wolf, G.

    1986-01-01

    We show that the large curvature in the T-P melting curve of KCl is the result of a reordering of the liquid to a more densely packed arrangement. As a result theories of melting, such as the instability model, which do not take into account the structure of the liquid fail to predict the correct pressure dependence of the melting curve

  8. Arctic sea ice melt pond fractal dimension - explained

    Science.gov (United States)

    Popovic, Predrag

    As Arctic sea ice starts to melt in the summer, pools of melt water quickly form on its surface, significantly changing its albedo, and impacting its subsequent evolution. These melt ponds often form complex geometric shapes. One characteristic of their shape, the fractal dimension of the pond boundaries, D, when plotted as a function of pond size, has been shown to transition between the two fundamental limits of D = 1 and D = 2 at some critical pond size. Here, we provide an explanation for this behavior. First, using aerial photographs, we show how this fractal transition curve changes with time, and show that there is a qualitative difference in the pond shape as ice transitions from impermeable to permeable. Namely, while ice is impermeable, maximum fractal dimension is less than 2, whereas after it becomes permeable, maximum fractal dimension becomes very close to 2. We then show how the fractal dimension of a collection of overlapping circles placed randomly on a plane also transitions from D = 1 to D = 2 at a size equal to the average size of a single circle. We, therefore, conclude that this transition is a simple geometric consequence of regular shapes connecting. The one physical parameter that can be extracted from the fractal transition curve is the length scale at which transition occurs. We provide a possible explanation for this length scale by noting that the flexural wavelength of the ice poses a fundamental limit on the size of melt ponds on permeable ice. If this is true, melt ponds could be used as a proxy for ice thickness.

  9. Permeability and 3-D melt geometry in shear-induced high melt fraction conduits

    Science.gov (United States)

    Zhu, W.; Cordonnier, B.; Qi, C.; Kohlstedt, D. L.

    2017-12-01

    Observations of dunite channels in ophiolites and uranium-series disequilibria in mid-ocean ridge basalt suggest that melt transport in the upper mantle beneath mid-ocean ridges is strongly channelized. Formation of high melt fraction conduits could result from mechanical shear, pyroxene dissolution, and lithological partitioning. Deformation experiments (e.g. Holtzman et al., 2003) demonstrate that shear stress causes initially homogeneously distributed melt to segregate into an array of melt-rich bands, flanked by melt-depleted regions. At the same average melt fraction, the permeability of high melt fraction conduits could be orders of magnitude higher than that of their homogenous counterparts. However, it is difficult to determine the permeability of melt-rich bands. Using X-ray synchrotron microtomography, we obtained high-resolution images of 3-dimensional (3-D) melt distribution in a partially molten rock containing shear-induced high melt fraction conduits. Sample CQ0705, an olivine-alkali basalt aggregate with a nominal melt fraction of 4%, was deformed in torsion at a temperature of 1473 K and a confining pressure of 300 MPa to a shear strain of 13.3. A sub-volume of CQ0705 encompassing 3-4 melt-rich bands was imaged. Microtomography data were reduced to binary form so that solid olivine is distinguishable from basalt glass. At a spatial resolution of 160 nm, the 3-D images reveal the shape and connectedness of melt pockets in the melt-rich bands. Thin melt channels formed at grain edges are connected at large melt nodes at grain corners. Initial data analysis shows a clear preferred orientation of melt pockets alignment subparallel to the melt-rich band. We use the experimentally determined geometrical parameters of melt topology to create a digital rock with identical 3-D microstructures. Stokes flow simulations are conducted on the digital rock to obtain the permeability tensor. Using this digital rock physics approach, we determine how deformation

  10. The modeling and analysis of in-vessel corium/structure interaction in boiling water reactors

    International Nuclear Information System (INIS)

    Podowski, M.Z.; Kurul, N.; Kim, S.-W.; Baltyn, W.; Frid, W.

    1997-01-01

    A complete stand-alone state-of-the-art model has been developed of the interaction between corium debris in the lower plenum and the RPV walls and internal structures, including the vessel failure mechanisms. This new model has been formulated as a set of consistent computer modules which could be linked with other existing models and/or computer codes. The combined lower head and lower plenum modules were parametrically tested and applied to predict the consequences of a hypothetical station blackout in a Swedish BWR. (author)

  11. The impact of microwave stray radiation to in-vessel diagnostic components

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, M.; Laqua, H. P.; Hathiramani, D.; Baldzuhn, J.; Biedermann, C.; Cardella, A.; Erckmann, V.; König, R.; Köppen, M.; Zhang, D. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, EURATOM Association, D-17489 Greifswald (Germany); Oosterbeek, J.; Brand, H. von der; Parquay, S. [Technische Universiteit Eindhoven, department Technische Natuurkunde, working group for Plasma Physics and Radiation Technology, Den Doelch 2, 5612 AZ Eindhoven (Netherlands); Jimenez, R. [Centro de Investigationes Energeticas, Medioambientales y Technológicas, Association EURATOM/CIEMAT, Avenida Complutense 22, Madrid 28040 (Spain); Collaboration: W7-X Teasm

    2014-08-21

    Microwave stray radiation resulting from unabsorbed multiple reflected ECRH / ECCD beams may cause severe heating of microwave absorbing in-vessel components such as gaskets, bellows, windows, ceramics and cable insulations. In view of long-pulse operation of WENDELSTEIN-7X the MIcrowave STray RAdiation Launch facility, MISTRAL, allows to test in-vessel components in the environment of isotropic 140 GHz microwave radiation at power load of up to 50 kW/m{sup 2} over 30 min. The results show that both, sufficient microwave shielding measures and cooling of all components are mandatory. If shielding/cooling measures of in-vessel diagnostic components are not efficient enough, the level of stray radiation may be (locally) reduced by dedicated absorbing ceramic coatings on cooled structures.

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

    International Nuclear Information System (INIS)

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

    1987-09-01

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

  13. Effect of Topography on Subglacial Discharge and Submarine Melting During Tidewater Glacier Retreat

    Science.gov (United States)

    Amundson, J. M.; Carroll, D.

    2018-01-01

    To first order, subglacial discharge depends on climate, which determines precipitation fluxes and glacier mass balance, and the rate of glacier volume change. For tidewater glaciers, large and rapid changes in glacier volume can occur independent of climate change due to strong glacier dynamic feedbacks. Using an idealized tidewater glacier model, we show that these feedbacks produce secular variations in subglacial discharge that are influenced by subglacial topography. Retreat along retrograde bed slopes (into deep water) results in rapid surface lowering and coincident increases in subglacial discharge. Consequently, submarine melting of glacier termini, which depends on subglacial discharge and ocean thermal forcing, also increases during retreat into deep water. Both subglacial discharge and submarine melting subsequently decrease as glacier termini retreat out of deep water and approach new steady state equilibria. In our simulations, subglacial discharge reached peaks that were 6-17% higher than preretreat values, with the highest values occurring during retreat from narrow sills, and submarine melting increased by 14% for unstratified fjords and 51% for highly stratified fjords. Our results therefore indicate that submarine melting acts in concert with iceberg calving to cause tidewater glacier termini to be unstable on retrograde beds. The full impact of submarine melting on tidewater glacier stability remains uncertain, however, due to poor understanding of the coupling between submarine melting and iceberg calving.

  14. Conceptual design studies of in-vessel viewing equipment for ITER

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Oka, Kiyoshi; Taguchi, Hiroshi; Itoh, Akira; Tada, Eisuke; Shibanuma, Kiyoshi

    1996-03-01

    In-vessel viewing systems are essential to inspect all surface of in-vessel components so as to detect and locate damages, and to assist in-vessel maintenance operations. The in-vessel viewing operations are categorized into the three cases, which are 1) rapid inspection just after off-normal events such as disruption, 2) scheduled inspection, and 3) supplementary inspection during maintenance operations. In case of the rapid inspection, the viewing systems have to be operated in vacuum (ca. 10 -5 Pa) and high temperature (ca. 300degC) under a gamma ray dose rate of 10 7 R/h. On the other hand, the latter two cases are anticipated to be under atmospheric inert gas, 150degC and 3x10 6 R/h. Accordingly, the in-vessel viewing systems are required to have sufficient durability under those conditions of all cases as well as precision of the vision to all of in-vessel surface. Based on those requirements, scoping studies on various viewing concepts have been performed and the applicability to the ITER conditions have been assessed. As a result, two types of viewing systems have been chosen, which are a periscope type viewing system and a image fiber type viewing system with a multi-joint manipulator. Both systems are based on radiation hard optical elements which are being developed. In this report, the design features of both viewing systems are described, including technical issues for ITER application. Finally, a periscope type viewing system is recommended as a primary system and the following specifications/conditions are proposed for the further engineering design. (1) Unified type periscope with a movable mirror at the tip (2) Integrated lighting device into the periscope (3) Accessed from top vertical ports located at 7.3m from the machine center (4) Proposed configuration with a total length of around 27m and a diameter of 200mm. (author)

  15. Electro-mechanical connection system for ITER in-vessel magnetic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Rizzolo, Andrea; Brombin, Matteo; Gonzalez, Winder [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Marconato, Nicolò, E-mail: nicolo.marconato@igi.cnr.it [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Peruzzo, Simone [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Arshad, Shakeib [Fusion for Energy, C/Josep Pla, 2, 08019 Barcelona (Spain); Ma, Yunxing; Vayakis, George [ITER Organization, Route de Vinon-sur-Verdon, 13067 St Paul Lez Durance (France); Williams, Adrian [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon, Oxon, OX14 1RL (United Kingdom)

    2016-11-01

    Highlights: • Latest status of the ITER “Generic In-Vessel Magnetic Platform” design activity. • Integration within the ITER In-Vessel configuration model. • Geometry optimization based on thermo-mechanical and magnetic field 3D calculation. • Assessment of the remote handling maintenance compatibility. - Abstract: This paper presents the preliminary design of the “In-Vessel Magnetic platform”, which is a subsystem of the magnetic diagnostics formed by all the components necessary for guaranteeing the thermo-mechanical interface of the actual magnetic sensors with the vacuum vessel (VV), their protection and the electrical connection to the in-vessel wiring for the transmission of the detected signal with a minimum level of noise. The design has been developed in order to comply with different functional requirements: the mechanical attachment to the VV; the electrical connection to the in-vessel wiring; efficient heat transfer to the VV; the compatibility with Remote Handling (RH) system for replacement; the integration of metrology features for post-installation control; the Electro Magnetic Interference (EMI) shielding from Electron Cyclotron Heating (ECH) stray radiation without compromising the sensor pass band (15 kHz). Significant effort has been dedicated to develop the CAD model, integrated within the ITER In-Vessel configuration model, taking care of the geometrical compliance with the Blanket modules (modified in order to accommodate the magnetic sensors in suitable grooves) and the RH compatibility. Thorough thermo-mechanical and electro-magnetic Finite Element Method (FEM) analyses have been performed to assess the reliability of the system in standard and off-normal operating conditions for the low frequency magnetic sensors.

  16. Recent Changes in the Arctic Melt Season

    Science.gov (United States)

    Stroeve, Julienne; Markus, Thorsten; Meier, Walter N.; Miller, Jeff

    2007-01-01

    Melt-season duration, melt-onset and freeze-up dates are derived from satellite passive microwave data and analyzed from 1979 to 2005 over Arctic sea ice. Results indicate a shift towards a longer melt season, particularly north of Alaska and Siberia, corresponding to large retreats of sea ice observed in these regions. Although there is large interannual and regional variability in the length of the melt season, the Arctic is experiencing an overall lengthening of the melt season at a rate of about 2 weeks decade(sup -1). In fact, all regions in the Arctic (except for the central Arctic) have statistically significant (at the 99% level or higher) longer melt seasons by greater than 1 week decade(sup -1). The central Arctic shows a statistically significant trend (at the 98% level) of 5.4 days decade(sup -1). In 2005 the Arctic experienced its longest melt season, corresponding with the least amount of sea ice since 1979 and the warmest temperatures since the 1880s. Overall, the length of the melt season is inversely correlated with the lack of sea ice seen in September north of Alaska and Siberia, with a mean correlation of -0.8.

  17. Niobium interaction with chloride-carbonate melts

    International Nuclear Information System (INIS)

    Kuznetsov, S.A.; Kuznetsova, S.V.

    1996-01-01

    Niobium interaction with chloride-carbonate melt NaCl-KCl-K 2 CO 3 (5 mass %) in the temperature range of 973-1123 K has been studied. The products and niobium corrosion rate have been ascertained, depending on the temperature of melt and time of allowance. Potentials of niobium corrosion have been measured. Refs. 11, figs. 3, tabs. 2

  18. Attenuation in Melting Layer of Precipitation

    NARCIS (Netherlands)

    Klaassen, W.

    1988-01-01

    A model of the melting layer is employed on radar measurements to simulate the attenuation of radio waves at 12, 20 and 30GHz. The attenuation in the melting layer is simulated to be slightly larger than that of rain with the same path length and precipitation intensity. The result appears to depend

  19. Multiscale approach to equilibrating model polymer melts

    DEFF Research Database (Denmark)

    Svaneborg, Carsten; Ali Karimi-Varzaneh, Hossein; Hojdis, Nils

    2016-01-01

    We present an effective and simple multiscale method for equilibrating Kremer Grest model polymer melts of varying stiffness. In our approach, we progressively equilibrate the melt structure above the tube scale, inside the tube and finally at the monomeric scale. We make use of models designed...

  20. Disordering and Melting of Aluminum Surfaces

    DEFF Research Database (Denmark)

    Stoltze, Per; Nørskov, Jens Kehlet; Landman, U.

    1988-01-01

    We report on a molecular-dynamics simulation of an Al(110) surface using the effective-medium theory to describe the interatomic interactions. The surface region is found to start melting ≅200 K below the bulk melting temperature with a gradual increase in the thickness of the disordered layer as...

  1. Water

    Science.gov (United States)

    ... drink and water in food (like fruits and vegetables). 6. Of all the earth’s water, how much is ocean or seas? 97 percent of the earth’s water is ocean or seas. 7. How much of the world’s water is frozen? Of all the water on earth, about 2 percent is frozen. 8. How much ...

  2. Mock-up tests of rail-mounted vehicle type in-vessel transporter/manipulator

    International Nuclear Information System (INIS)

    Oka, K.; Kakaudate, S.; Fukatsu, S.

    1995-01-01

    A rail-mounted vehicle system has been developed for remote maintenance of in-vessel components for fusion experimental reactor. In this system, a rail deploying/storing system is installed at outside of the reactor core and used to deploy a rail transporter and vehicle/manipulator for the in-vessel maintenance. A prototype of the rail deploying/storing system has been fabricated for mockup tests. This paper describes structural design of the prototypical rail deploying/storing system and results of the performance tests such as payload capacity, position control and rail deployment/storage performance

  3. Shielding analysis of the LMR in-vessel fuel storage experiments

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    1994-01-01

    The In-Vessel Fuel Storage (IVFS) experiments analyzed in this paper were conducted at the Oak Ridge National Laboratory's Tower Shielding Reactor (TSR) as part of the Japanese-American Shielding Program for Experimental Research (JASPER). These IVFS experiments were designed to study source multiplication and three-dimensional effects related to in-vessel storage of spent fuel elements in liquid metal reactor (LMR) systems. The present paper describes the 2- and 3-D calculations and results corresponding to a limited subset of those IVFS experiments in which the US LMR program had a particular interest

  4. Design system for in-vessel mainipulator of fusion reactor 'DESIM'

    International Nuclear Information System (INIS)

    Adachi, Junihci; Kobayashi, Takeshi; Ise, Hideo; Sato, Keisuke; Matsuda, Hirotsugu

    1989-01-01

    A computer aided design system 'DESIM' for the in-vessel manipulators of nuclear fusion reactors has been developed to design the manipulators efficiently. The DESIM consists of the following subsystems: (1) the design system for arm mechanisms to realize optimum manipulation performance in the specified workspace; (2) the robot simulator to study manipulator movement, postures and interference problems; (3) the CAD system which is used to define the structure object data for robots, and the interface system for the data conversion from the CAD system to the robot simulator. The DESIM has been used to design the in-vessel manipulator for the Fusion Experimental Reactor (FER) to confirm the effectiveness. (author)

  5. In-vessel tritium retention and removal in ITER

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-01

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

  6. In-vessel tritium retention and removal in ITER

    International Nuclear Information System (INIS)

    Federici, G.; Anderl, R.A.

    1998-01-01

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

  7. Shape evolution of a melting nonspherical particle

    Science.gov (United States)

    Kintea, Daniel M.; Hauk, Tobias; Roisman, Ilia V.; Tropea, Cameron

    2015-09-01

    In this study melting of irregular ice crystals was observed in an acoustic levitator. The evolution of the particle shape is captured using a high-speed video system. Several typical phenomena have been discovered: change of the particle shape, appearance of a capillary flow of the melted liquid on the particle surface leading to liquid collection at the particle midsection (where the interface curvature is smallest), and appearance of sharp cusps at the particle tips. No such phenomena can be observed during melting of spherical particles. An approximate theoretical model is developed which accounts for the main physical phenomena associated with melting of an irregular particle. The agreement between the theoretical predictions for the melting time, for the evolution of the particle shape, and the corresponding experimental data is rather good.

  8. Nanotexturing of surfaces to reduce melting point.

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, Ernest J.; Zubia, David (University of Texas at El Paso El Paso, TX); Mireles, Jose (Universidad Aut%C3%94onoma de Ciudad Ju%C3%94arez Ciudad Ju%C3%94arez, Mexico); Marquez, Noel (University of Texas at El Paso El Paso, TX); Quinones, Stella (University of Texas at El Paso El Paso, TX)

    2011-11-01

    This investigation examined the use of nano-patterned structures on Silicon-on-Insulator (SOI) material to reduce the bulk material melting point (1414 C). It has been found that sharp-tipped and other similar structures have a propensity to move to the lower energy states of spherical structures and as a result exhibit lower melting points than the bulk material. Such a reduction of the melting point would offer a number of interesting opportunities for bonding in microsystems packaging applications. Nano patterning process capabilities were developed to create the required structures for the investigation. One of the technical challenges of the project was understanding and creating the specialized conditions required to observe the melting and reshaping phenomena. Through systematic experimentation and review of the literature these conditions were determined and used to conduct phase change experiments. Melting temperatures as low as 1030 C were observed.

  9. Melting Can Hinder Impact-Induced Adhesion

    Science.gov (United States)

    Hassani-Gangaraj, Mostafa; Veysset, David; Nelson, Keith A.; Schuh, Christopher A.

    2017-10-01

    Melting has long been used to join metallic materials, from welding to selective laser melting in additive manufacturing. In the same school of thought, localized melting has been generally perceived as an advantage, if not the main mechanism, for the adhesion of metallic microparticles to substrates during a supersonic impact. Here, we conduct the first in situ supersonic impact observations of individual metallic microparticles aimed at the explicit study of melting effects. Counterintuitively, we find that under at least some conditions melting is disadvantageous and hinders impact-induced adhesion. In the parameter space explored, i.e., ˜10 μ m particle size and ˜1 km /s particle velocity, we argue that the solidification time is much longer than the residence time of the particle on the substrate, so that resolidification cannot be a significant factor in adhesion.

  10. Modeling the summertime evolution of sea-ice melt ponds

    DEFF Research Database (Denmark)

    Lüthje, Mikael; Feltham, D.L.; Taylor, P.D.

    2006-01-01

    We present a mathematical model describing the summer melting of sea ice. We simulate the evolution of melt ponds and determine area coverage and total surface ablation. The model predictions are tested for sensitivity to the melt rate of unponded ice, enhanced melt rate beneath the melt ponds...

  11. A model for the computation of the thermal processes in the reactor cavity during a severe accident in a LWR, at the presence of sump water, from the time of reactor pressure vessel failure to the start time of melt/concrete interaction

    International Nuclear Information System (INIS)

    Hirschmann, H.

    1990-04-01

    At present no experimental results are available which analyze that stage of a severe accident in a light water reactor, during which the reactor pressure vessel fails by melting, the core debris relocates into the water pool on the floor of the containment building (cavity) and again is heated up. Therefore an analytical model is described, with the help of which the process of material relocation, the heating of the material in the cavity interacting with the pool water, and the production rates of vapour and hydrogen can be estimated. The slumped mass accumulating in the cavity is taken to be the sum of infinitely small mass parts, assumed to slump at different times, which after slumping undergo individual thermal histories. The enthalpy of the slumped mass is the sum of the enthalpies of the single mass parts. The average temperature of the slumped mass is given by the enthalpy computed in this manner. The production rates of the gases are additive superpositions of all partial rates from the mass parts. The gas rates are computed using the balance of enthalpy and mass. (author) 5 refs

  12. Deep pooling of low degree melts and volatile fluxes at the 85°E segment of the Gakkel Ridge: Evidence from olivine-hosted melt inclusions and glasses

    Science.gov (United States)

    Shaw, Alison M.; Behn, Mark D.; Humphris, Susan E.; Sohn, Robert A.; Gregg, Patricia M.

    2010-01-01

    We present new analyses of volatile, major, and trace elements for a suite of glasses and melt inclusions from the 85°E segment of the ultra-slow spreading Gakkel Ridge. Samples from this segment include limu o pele and glass shards, proposed to result from CO 2-driven explosive activity. The major element and volatile compositions of the melt inclusions are more variable and consistently more primitive than the glass data. CO 2 contents in the melt inclusions extend to higher values (167-1596 ppm) than in the co-existing glasses (187-227 ppm), indicating that the melt inclusions were trapped at greater depths. These melt inclusions record the highest CO 2 melt concentrations observed for a ridge environment. Based on a vapor saturation model, we estimate that the melt inclusions were trapped between seafloor depths (˜ 4 km) and ˜ 9 km below the seafloor. However, the glasses are all in equilibrium with their eruption depths, which is inconsistent with the rapid magma ascent rates expected for explosive activity. Melting conditions inferred from thermobarometry suggest relatively deep (25-40 km) and cold (1240°-1325 °C) melting conditions, consistent with a thermal structure calculated for the Gakkel Ridge. The water contents and trace element compositions of the melt inclusions and glasses are remarkably homogeneous; this is an unexpected result for ultra-slow spreading ridges, where magma mixing is generally thought to be less efficient based on the assumption that steady-state crustal magma chambers are absent in these environments. All melts can be described by a single liquid line of descent originating from a pooled melt composition that is consistent with the aggregate melt calculated from a geodynamic model for the Gakkel Ridge. These data suggest a model in which deep, low degree melts are efficiently pooled in the upper mantle (9-20 km depth), after which crystallization commences and continues during ascent and eruption. Based on our melting model

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

    International Nuclear Information System (INIS)

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

    1999-03-01

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

  14. The Impact Of Snow Melt On Surface Runoff Of Sava River In Slovenia

    Science.gov (United States)

    Horvat, A.; Brilly, M.; Vidmar, A.; Kobold, M.

    2009-04-01

    Snow is a type of precipitation in the form of crystalline water ice, consisting of a multitude of snowflakes that fall from clouds. Snow remains on the ground until it melts or sublimates. Spring snow melt is a major source of water supply to areas in temperate zones near mountains that catch and hold winter snow, especially those with a prolonged dry summer. In such places, water equivalent is of great interest to water managers wishing to predict spring runoff and the water supply of cities downstream. In temperate zone like in Slovenia the snow melts in the spring and contributes certain amount of water to surface flow. This amount of water can be great and can cause serious floods in case of fast snow melt. For this reason we tried to determine the influence of snow melt on the largest river basin in Slovenia - Sava River basin, on surface runoff. We would like to find out if snow melt in Slovenian Alps can cause spring floods and how serious it can be. First of all we studied the caracteristics of Sava River basin - geology, hydrology, clima, relief and snow conditions in details for each subbasin. Furtermore we focused on snow and described the snow phenomenom in Slovenia, detailed on Sava River basin. We collected all available data on snow - snow water equivalent and snow depth. Snow water equivalent is a much more useful measurement to hydrologists than snow depth, as the density of cool freshly fallen snow widely varies. New snow commonly has a density of between 5% and 15% of water. But unfortunately there is not a lot of available data of SWE available for Slovenia. Later on we compared the data of snow depth and river runoff for some of the 40 winter seasons. Finally we analyzed the use of satellite images for Slovenia to determine the snow cover for hydrology reason. We concluded that snow melt in Slovenia does not have a greater influence on Sava River flow. The snow cover in Alps can melt fast due to higher temperatures but the water distributes

  15. Water

    International Nuclear Information System (INIS)

    Chovanec, A.; Grath, J.; Kralik, M.; Vogel, W.

    2002-01-01

    An up-date overview of the situation of the Austrian waters is given by analyzing the status of the water quality (groundwater, surface waters) and water protection measures. Maps containing information of nitrate and atrazine in groundwaters (analyses at monitoring stations), nitrate contents and biological water quality of running waters are included. Finally, pollutants (nitrate, orthophosphate, ammonium, nitrite, atrazine etc.) trends in annual mean values and median values for the whole country for the years 1992-1999 are presented in tables. Figs. 5. (nevyjel)

  16. The COMET-L3 experiment on long-term melt. Concrete interaction and cooling by surface flooding

    International Nuclear Information System (INIS)

    Alsmeyer, H.; Cron, T.; Fluhrer, B.; Messemer, G.; Miassoedov, A.; Schmidt-Stiefel, S.; Wenz, T.

    2007-02-01

    The COMET-L3 experiment considers the long-term situation of corium/concrete interaction in an anticipated core melt accident of a light-water-reactor, after the metal melt is layered beneath the oxide melt. The experimental focus is on cavity formation in the basemat and the risk of long term basemat penetration. The experiment investigates the two-dimensional concrete erosion in a cylindrical crucible fabricated from siliceous concrete in the first phase of the test, and the influence of surface flooding in the second phase. Decay heating in the two-component metal and oxide melt is simulated by sustained induction heating of the metal phase that is overlaid by the oxide melt. The inner diameter of the concrete crucible was 60 cm, the initial mass of the melt was 425 kg steel and 211 kg oxide at 1665 C, resulting in a melt height of 450 mm. The net power to the metal melt was about 220 kW from 0 s to 1880 s, when the maximum erosion limit of the crucible was reached and heating was terminated. In the initial phase of the test (less than 100 s), the overheated, highly agitated metal melt causes intense interaction with the concrete, which leads to fast decrease of the initial melt overheat and reduction of the initially high concrete erosion rate. Thereafter, under quasistationary conditions until about 800 s, the erosion by the metal melt slows down to some 0.07 mm/s into the axial direction. Lateral erosion is a factor 3 smaller. Video observation of the melt surface shows an agitated melt with ongoing gas release from the decomposing concrete. Several periods of more intense gas release, gas driven splashing, and release of crusts from the concrete interface indicate the existence and iterative break-up of crusts that probably form at the steel/concrete interface. Surface flooding of the melt is initiated at 800 s by a shower from the crucible head with 0.375 litre water/s. Flooding does not lead to strong melt/water interactions, and no entrapment reactions or

  17. Effects of Melt Processing on Evolution of Structure in PEEK

    Science.gov (United States)

    Georgiev, Georgi; Dai, Patrick Shuanghua; Oyebode, Elizabeth; Cebe, Peggy; Capel, Malcolm

    1999-01-01

    treatment scheme involving annealing/crystallization at T(sub a1) followed by annealing at T(sub a2) where either T(sub a1) T(sub a2). We proposed a model to explain multiple melting endotherms in PPS, treated according to one or two-stage melt or cold crystallization. Key features of this model are that multiple endotherms: (1) are due to reorganization/recrystallization after cold crystallization; and, (2) are dominated by crystal morphology after melt crystallization at high T. In other words, multiple distinct crystal populations are formed by the latter treatment, leading to observation of multiple melting. PEEK 45OG pellets (ICI Americas) were the starting material for this study. Films were compression molded at 400 C, then quenched to ice water. Samples were heated to 375 C in a Mettler FP80 hot stage and held for three min. to erase crystal seeds before cooling them to T(sub a1) = 280 C . Samples were held at T(sub a2) for a period of time, then immediately heated to 360 C. In the second treatment samples were held at T(sub a1) = 31 C for different crystallization times t(sub c) then cooled to 295 C and held 15 min. In situ (SAXS) experiments were performed at the Brookhaven National Synchrotron Light Source with the sample located inside the Mettler hot stage. The system was equipped with a two-dimensional position sensitive detector. The sample to detector distance was 172.7 cm and the X-ray wavelength was 1.54 Angstroms. SAXS data were taken continuously during the isothermal periods and during the heating to 360 C at 5 C/min. Each SAXS scan was collected for 30 sec. Since the samples were isotropic, circular integration was used to increase the signal to noise ratio. After dual stage melt crystallization with T(sub a1) T(sub a2), the amount of space remaining for additional growth at T(sub a2) depends upon the holding time at T(sub a1). The long period of crystals formed at T(sub a2) is smaller than that formed at T(sub a1) due to growth in a now

  18. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  19. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    International Nuclear Information System (INIS)

    Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

    2015-01-01

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  20. Viscosity of ring polymer melts

    KAUST Repository

    Pasquino, Rossana

    2013-10-15

    We have measured the linear rheology of critically purified ring polyisoprenes, polystyrenes, and polyethyleneoxides of different molar masses. The ratio of the zero-shear viscosities of linear polymer melts η0,linear to their ring counterparts η0,ring at isofrictional conditions is discussed as a function of the number of entanglements Z. In the unentangled regime η0,linear/η 0,ring is virtually constant, consistent with the earlier data, atomistic simulations, and the theoretical expectation η0,linear/ η0,ring = 2. In the entanglement regime, the Z-dependence of ring viscosity is much weaker than that of linear polymers, in qualitative agreement with predictions from scaling theory and simulations. The power-law extracted from the available experimental data in the rather limited range 1 < Z < 20, η0,linear/η0,ring ∼ Z 1.2±0.3, is weaker than the scaling prediction (η0,linear/η0,ring ∼ Z 1.6±0.3) and the simulations (η0,linear/ η0,ring ∼ Z2.0±0.3). Nevertheless, the present collection of state-of-the-art experimental data unambiguously demonstrates that rings exhibit a universal trend clearly departing from that of their linear counterparts, and hence it represents a major step toward resolving a 30-year-old problem. © 2013 American Chemical Society.

  1. Viscosity of ring polymer melts

    KAUST Repository

    Pasquino, Rossana; Vasilakopoulos, Thodoris C.; Jeong, Youncheol; Lee, Hyojoon; Rogers, Simon A.; Sakellariou, Georgios; Allgaier, Jü rgen B.; Takano, Atsushi; Brá s, Ana Rita E; Chang, Taihyun; Gooß en, Sebastian; Pyckhout-Hintzen, Wim; Wischnewski, Andreas; Hadjichristidis, Nikolaos; Richter, Dieter R.; Rubinstein, Michael H.; Vlassopoulos, Dimitris

    2013-01-01

    We have measured the linear rheology of critically purified ring polyisoprenes, polystyrenes, and polyethyleneoxides of different molar masses. The ratio of the zero-shear viscosities of linear polymer melts η0,linear to their ring counterparts η0,ring at isofrictional conditions is discussed as a function of the number of entanglements Z. In the unentangled regime η0,linear/η 0,ring is virtually constant, consistent with the earlier data, atomistic simulations, and the theoretical expectation η0,linear/ η0,ring = 2. In the entanglement regime, the Z-dependence of ring viscosity is much weaker than that of linear polymers, in qualitative agreement with predictions from scaling theory and simulations. The power-law extracted from the available experimental data in the rather limited range 1 < Z < 20, η0,linear/η0,ring ∼ Z 1.2±0.3, is weaker than the scaling prediction (η0,linear/η0,ring ∼ Z 1.6±0.3) and the simulations (η0,linear/ η0,ring ∼ Z2.0±0.3). Nevertheless, the present collection of state-of-the-art experimental data unambiguously demonstrates that rings exhibit a universal trend clearly departing from that of their linear counterparts, and hence it represents a major step toward resolving a 30-year-old problem. © 2013 American Chemical Society.

  2. Cladding hull decontamination and densification process. Part 2. Densification by inductoslag melting

    International Nuclear Information System (INIS)

    Nelson, R.G.; Montgomery, D.R.

    1980-04-01

    The Inductoslag melting process was developed to densify Zircaloy-4 cladding hulls. It is a cold crucible process that uses induction heating, a segmented water-cooled copper crucible, and a calcium fluoride flux. Metal and flux are fed into the furnace through the crucible, located at the top of the furnace, and the finished ingot is withdrawn from the bottom of the furnace. Melting rates of 40 to 50 kg/h are achieved, using 100 to 110 kW at an average energy use of 2.5 kWh/kg. The quality of ingots produced from factory supplied cladding tubing is sufficient to satisfy nuclear grade standards. An ingot of Zircaloy-4, made from melted cladding tubing that had been autoclaved to near reactor exposure and then descaled by the hydrogen fluoride decontamination process prior to Inductoslag melting, did not meet nuclear grade standards because the hydrogen, nitrogen, and hardness levels were too high. Melting development work is described that could possibly be used to test the capability of the Inductoslag process to satisfactorily melt a variety and mix of materials from LWR reprocessing, decontamination, and storage options. Results of experiments are also presented that could be used to improve remote operation of the melting process

  3. Water

    Science.gov (United States)

    ... can be found in some metal water taps, interior water pipes, or pipes connecting a house to ... reduce or eliminate lead. See resources below. 5. Children and pregnant women are especially vulnerable to the ...

  4. Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

    International Nuclear Information System (INIS)

    Parker, R.R.

    1998-01-01

    This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

  5. Dynamics of Melting and Melt Migration as Inferred from Incompatible Trace Element Abundance in Abyssal Peridotites

    Science.gov (United States)

    Peng, Q.; Liang, Y.

    2008-12-01

    To better understand the melting processes beneath the mid-ocean ridge, we developed a simple model for trace element fractionation during concurrent melting and melt migration in an upwelling steady-state mantle column. Based on petrologic considerations, we divided the upwelling mantle into two regions: a double- lithology upper region where high permeability dunite channels are embedded in a lherzolite/harzburgite matrix, and a single-lithology lower region that consists of partially molten lherzolite. Melt generated in the single lithology region migrates upward through grain-scale diffuse porous flow, whereas melt in the lherzolite/harzburgite matrix in the double-lithology region is allowed to flow both vertically through the overlying matrix and horizontally into its neighboring dunite channels. There are three key dynamic parameters in our model: degree of melting experienced by the single lithology column (Fd), degree of melting experienced by the double lithology column (F), and a dimensionless melt suction rate (R) that measures the accumulated rate of melt extraction from the matrix to the channel relative to the accumulated rate of matrix melting. In terms of trace element fractionation, upwelling and melting in the single lithology column is equivalent to non-modal batch melting (R = 0), whereas melting and melt migration in the double lithology region is equivalent to a nonlinear combination of non-modal batch and fractional melting (0 abyssal peridotite, we showed, with the help of Monte Carlo simulations, that it is difficult to invert for all three dynamic parameters from a set of incompatible trace element data with confidence. However, given Fd, it is quite possible to constrain F and R from incompatible trace element abundances in residual peridotite. As an illustrative example, we used the simple melting model developed in this study and selected REE and Y abundance in diopside from abyssal peridotites to infer their melting and melt migration

  6. The corrosion of steels by hot sodium melts

    International Nuclear Information System (INIS)

    Currie, R.

    1996-01-01

    Considerable research has been performed by AEA Technology on the corrosion of steels by hot sodium melts containing sodium hydroxide and sodium oxide. This research has principally been in support of understanding the effects of sodium-water reactions on the internals of fast reactor steam generators. The results however have relevance to sodium fires. It has been determined that the rate of corrosion of steels by melts of pure NaOH can be significantly increased by the addition of Na 2 O. In the case of a sodium-water reaction jet created by a leak of steam into sodium, the composition of the jet varies from 100% sodium through to 100% steam, with a full range of concentrations of NaOH and Na 2 O, depending on axial and radial position. The temperature in the jet also varies with position, ranging from bulk sodium temperature on one boundary to expanded steam temperature on the other boundary, with internal temperatures ranging up to 1300 deg. C, depending on the local pre-reaction mole ratio of steam to sodium. In the case of sodium-water reaction jets, it has been possible to develop a model which predicts the composition of the reaction jet and then, using the data generated on the corrosivity of sodium melts, predict the rate of corrosion of a steel target in the path of the jet. In the case of a spray sodium fire, the sodium will initially contain a concentration of NaOH and the combustion process will generate Na 2 O. If there is sufficient humidity, conversion of some of the Na 2 O to NaOH will also occur. There is therefore the potential for aggressive mixtures of NaOH and Na 2 O to exist on the surface of the sodium droplets. It is therefore possible that the rate of corrosion of steels in the path of the spray may be higher than expected on the basis of assuming that only Na and Na 2 O were present. In the case of a pool sodium fire, potentially corrosive mixtures of NaOH and Na 2 O may be formed at some locations on the surface. This could lead to

  7. Technical meeting on materials for in-vessel components of ITER

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.

    2000-01-01

    The Technical meeting on materials for in-vessel components of ITER was held at the ITER Joint Work Site in Garching from 31 January to 4 February. The main objectives of the meetings were: 1. to summarize the requirements, 2. to review new data, 3. to discuss in detail the R and D program and to discuss the material assessment report

  8. Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

  9. Complete dissipation of 2,4,6-trinitrotoluene by in-vessel composting

    NARCIS (Netherlands)

    Gümüscü, B.; Cekmecelioglu, Deniz; Tekinay, Turgay

    2015-01-01

    We demonstrate complete removal of 2,4,6-trinitrotoluene (TNT) in 15 days using an in-vessel composting system, which is amended with TNT-degrading bacteria strains. A mixture of TNT, food waste, manure, wood chips, soil and TNT-degrading bacteria consortium are co-composted for 15 days in an

  10. Progress on the design development and prototype manufacturing of the ITER In-vessel coils

    NARCIS (Netherlands)

    Encheva, A.; Omran, H.; Devred, A.; Vostner, A.; Mitchell, N.; Mariani, N.; Jun, CH H.; Long, F.; Zhou, C.; Macklin, B.; Marti, H. P.; Sborchia, C.; della Corte, A. Della; Di Zenobio, A.; Anemona, A.; Righetti, R.; Wu, Y.; Jin, H.; Xu, A.; Jin, J.

    2017-01-01

    ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements.

  11. 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)

  12. Corium melt researches at VESTA test facility

    Directory of Open Access Journals (Sweden)

    Hwan Yeol Kim

    2017-10-01

    Full Text Available VESTA (Verification of Ex-vessel corium STAbilization and VESTA-S (-small test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging ZrO2 melt jet on a sacrificial material were performed to investigate the ablation characteristics. ZrO2 melt in an amount of 65–70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40, and the other is a stainless steel (SUS304 melt. Metallic melt in an amount of 1.5–2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. ZrO2 melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is UO2 60%, Zr 10%, ZrO2 15%, SUS304 14%, and B4C 1%, was melted in a

  13. Melting technique for vanadium containing steels

    Energy Technology Data Exchange (ETDEWEB)

    Grishanov, M P; Gutovskij, I B; Vakhrushev, A S

    1980-04-28

    To descrease cost price of high-quality vanadium steels a method of their melting in open-hearth furnaces with acid lining using slag-metal fraction of vanadium, which is loaded in the content of 2.1-4.7% of melting mass, is suggested. Introduction of slag-metal fraction of vanadium ensures the formation of slag with composition that guarantees the necessary content of vanadium in steel and does not require introduction of expensive vanadium-containing ferroalloys into the melt.

  14. Melt processed high-temperature superconductors

    CERN Document Server

    1993-01-01

    The achievement of large critical currents is critical to the applications of high-temperature superconductors. Recent developments have shown that melt processing is suitable for producing high J c oxide superconductors. Using magnetic forces between such high J c oxide superconductors and magnets, a person could be levitated.This book has grown largely out of research works on melt processing of high-temperature superconductors conducted at ISTEC Superconductivity Research Laboratory. The chapters build on melt processing, microstructural characterization, fundamentals of flux pinning, criti

  15. Technological properties and structure of titanate melts

    International Nuclear Information System (INIS)

    Morozov, A.A.

    2002-01-01

    Power substantiation of existence of tough stream of complex anion ([TiO 6 ] 8- ) as a prevalent unit in titanate melts is given on the base of up-to-date knowledge about structure of metallurgical slags and results of investigations of thermophysical properties of these melts. It is shown that high crystallization ability of titanate melts at technological temperatures is determined by heterogeneity of liquid state - by presence up to 30 % of dispersed particles of solid phase solutions in matrix liquid [ru

  16. Bayesian estimation of core-melt probability

    International Nuclear Information System (INIS)

    Lewis, H.W.

    1984-01-01

    A very simple application of the canonical Bayesian algorithm is made to the problem of estimation of the probability of core melt in a commercial power reactor. An approximation to the results of the Rasmussen study on reactor safety is used as the prior distribution, and the observation that there has been no core melt yet is used as the single experiment. The result is a substantial decrease in the mean probability of core melt--factors of 2 to 4 for reasonable choices of parameters. The purpose is to illustrate the procedure, not to argue for the decrease

  17. Bubble Formation in Basalt-like Melts

    DEFF Research Database (Denmark)

    Jensen, Martin; Keding, Ralf; Yue, Yuanzheng

    2011-01-01

    and their diameter. The variation in melting temperature has little influence on the overall bubble volume. However, the size distribution of the bubbles varies with the melting temperature. When the melt is slowly cooled, the bubble volume increases, implying decreased solubility of the gaseous species. Mass...... spectroscopy analysis of gases liberated during heating of the glass reveals that small bubbles contain predominantly CH4, CO and CO2, whereas large bubbles bear N2, SO2 and H2S. The methodology utilised in this work can, besides mapping the bubbles in a glass, be applied to shed light on the sources of bubble...

  18. Slab melting beneath the Cascades Arc driven by dehydration of altered oceanic peridotite

    Science.gov (United States)

    Walowski, Kristina J; Wallace, Paul J.; Hauri, E.H.; Wada, I.; Clynne, Michael A.

    2015-01-01

    Water is returned to Earth’s interior at subduction zones. However, the processes and pathways by which water leaves the subducting plate and causes melting beneath volcanic arcs are complex; the source of the water—subducting sediment, altered oceanic crust, or hydrated mantle in the downgoing plate—is debated; and the role of slab temperature is unclear. Here we analyse the hydrogen-isotope and trace-element signature of melt inclusions in ash samples from the Cascade Arc, where young, hot lithosphere subducts. Comparing these data with published analyses, we find that fluids in the Cascade magmas are sourced from deeper parts of the subducting slab—hydrated mantle peridotite in the slab interior—compared with fluids in magmas from the Marianas Arc, where older, colder lithosphere subducts. We use geodynamic modelling to show that, in the hotter subduction zone, the upper crust of the subducting slab rapidly dehydrates at shallow depths. With continued subduction, fluids released from the deeper plate interior migrate into the dehydrated parts, causing those to melt. These melts in turn migrate into the overlying mantle wedge, where they trigger further melting. Our results provide a physical model to explain melting of the subducted plate and mass transfer from the slab to the mantle beneath arcs where relatively young oceanic lithosphere is subducted.

  19. Ocean stratification reduces melt rates at the grounding zone of the Ross Ice Shelf

    Science.gov (United States)

    Begeman, C. B.; Tulaczyk, S. M.; Marsh, O.; Mikucki, J.; Stanton, T. P.; Hodson, T. O.; Siegfried, M. R.; Powell, R. D.; Christianson, K. A.; King, M. A.

    2017-12-01

    Ocean-driven melting of ice shelves is often invoked as the primary mechanism for triggering ice loss from Antarctica. However, due to the difficulty in accessing the sub-ice-shelf ocean cavity, the relationship between ice-shelf melt rates and ocean conditions is poorly understood, particularly near the transition from grounded to floating ice, known as the grounding zone. Here we present the first borehole oceanographic observations from the grounding zone of Antarctica's largest ice shelf. Contrary to predictions that tidal currents near grounding zones should mix the water column, driving high ice-shelf melt rates, we find a stratified sub-ice-shelf water column. The vertical salinity gradient dominates stratification over a weakly unstable vertical temperature gradient; thus, stratification takes the form of a double-diffusive staircase. These conditions limit vertical heat fluxes and lead to low melt rates in the ice-shelf grounding zone. While modern grounding zone melt rates may presently be overestimated in models that assume efficient tidal mixing, the high sensitivity of double-diffusive staircases to ocean freshening and warming suggests future melt rates may be underestimated, biasing projections of global sea-level rise.

  20. Simulation of snow distribution and melt under cloudy conditions in an Alpine watershed

    Directory of Open Access Journals (Sweden)

    H.-Y. Li

    2011-07-01

    Full Text Available An energy balance method and remote-sensing data were used to simulate snow distribution and melt in an alpine watershed in northwestern China within a complete snow accumulation-melt period. The spatial energy budgets were simulated using meteorological observations and a digital elevation model of the watershed. A linear interpolation method was used to estimate the daily snow cover area under cloudy conditions, using Moderate Resolution Imaging Spectroradiometer (MODIS data. Hourly snow distribution and melt, snow cover extent and daily discharge were included in the simulated results. The root mean square error between the measured snow-water equivalent samplings and the simulated results is 3.2 cm. The Nash and Sutcliffe efficiency statistic (NSE between the measured and simulated discharges is 0.673, and the volume difference (Dv is 3.9 %. Using the method introduced in this article, modelling spatial snow distribution and melt runoff will become relatively convenient.

  1. Influence of gas generation on high-temperature melt/concrete interactions

    International Nuclear Information System (INIS)

    Powers, D.A.

    1979-01-01

    Accidents involving fuel melting and eventual contact between the high temperature melt and structural concrete may be hypothesized for both light water thermal reactors and liquid metal cooled breeder reactors. Though these hypothesized accidents have a quite low probability of occurring, it is necessary to investigate the probable natures of the accidents if an adequate assessment of the risks associated with the use of nuclear reactors is to be made. A brief description is given of a program addressing the nature of melt/concrete interactions which has been underway for three years at Sandia Laboratories. Emphasis in this program has been toward the behavior of prototypic melts of molten core materials with concrete representative of that found in existing or proposed reactors. The goals of the experimentation have been to identify phenomena particularly pertinent to questions of reactor safety, and phenomena particularly pertinent to questions of reactor safety, and provide quantitative data suitable for the purposes of risk assessment

  2. Consolidation of simulated nuclear metallic waste by vacuum coreless induction melting

    International Nuclear Information System (INIS)

    Montgomery, D.R.

    1984-10-01

    Vacuum coreless induction melting with bottom pouring has exceeded expectations for simplicity, reliability, and versatility when melting the zirconium and iron eutectic alloy. The melting tests have established that: the eutectic mixture of oxidized Zircaloy 4 hulls mixed with Type 316 stainless steel hulls can be melted at 41 kg/h at 40 kW with a power consumption of 1.03 kWh/kg and a melting temperature of 1260 0 C; the life of a graphite crucible can be expected to be longer by a factor of 4 than was previously projected; the bottom-pour water-cooled copper freeze plug was 100% reliable; a 24-in.-tall stainless steel canister with 1/4-in.-thick walls (6-in. inside diameter) was satisfactory in every respect; an ingot formed from 4 consecutive heats poured into a stainless steel canister appeared to be approx. 99% dense after sectioning; preplaced scrap in the canister can be encapsulated with molten metal to about 99% density; large pieces of Zircaloy 4 and stainless steel scrap can be melted, but have differing melting parameters; the pouring nozzle requires further development to prevent solidified drops from forming at the hole exit after a pour. It is recommended that a large-scale cold mock-up facility be established to refine and test a full-scale vacuum coreless induction melting system. Other options might include further scaled-down experiments to test other alloys and crucible materials under different atmospheric conditions (i.e., air melting). 1 reference, 18 figures, 1 table

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  4. The recycling through melting machining chips: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Luiz A.T.; Rossi, Jesualdo L., E-mail: luiz.atp@uol.com.br, E-mail: jelrossi@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil). Centro de Ciência e Tecnologia de Materiais

    2017-07-01

    PWR (Pressurized Water Reactor) reactors employ as nuclear fuel UO{sub 2} pellets packed in zirconium alloy tubes, called cladding. In the manufacture of the tubes, machining chips are generated which cannot be discarded, since the recycling of this material is strategic in terms of nuclear technology, legislation, economics and the environment. These nuclear alloys are very expensive and are not produced in Brazil and are imported for the manufacture of nuclear fuel. In this work, it will examined methods not yet studied to recycle Zircaloy chips using electron beam furnace in order to obtain ingots. In addition, it is intended to carry out the melting of new Zircaloy alloys, from the melting of zirconium sponge obtained in IPEN and imported and Zircaloy bars. The mechanical properties and the present phases of the material should be determined, as well as, the characterization of the microstructures by optical microscopy. This work, therefore, aims at the creation of a new line of research where methods will be approached to recycle the Zircaloy chips and to reduce in 30 times the volume by means of melting the enormous amount of material stored in the form of machining chips, being able to do others components for nuclear or chemical industry use, as well as conducting basic development research. (author)

  5. Observations of brine plumes below melting Arctic sea ice

    Directory of Open Access Journals (Sweden)

    A. K. Peterson

    2018-02-01

    Full Text Available In sea ice, interconnected pockets and channels of brine are surrounded by fresh ice. Over time, brine is lost by gravity drainage and flushing. The timing of salt release and its interaction with the underlying water can impact subsequent sea ice melt. Turbulence measurements 1 m below melting sea ice north of Svalbard reveal anticorrelated heat and salt fluxes. From the observations, 131 salty plumes descending from the warm sea ice are identified, confirming previous observations from a Svalbard fjord. The plumes are likely triggered by oceanic heat through bottom melt. Calculated over a composite plume, oceanic heat and salt fluxes during the plumes account for 6 and 9 % of the total fluxes, respectively, while only lasting in total 0.5 % of the time. The observed salt flux accumulates to 7.6 kg m−2, indicating nearly full desalination of the ice. Bulk salinity reduction between two nearby ice cores agrees with accumulated salt fluxes to within a factor of 2. The increasing fraction of younger, more saline ice in the Arctic suggests an increase in desalination processes with the transition to the new Arctic.

  6. Observations of brine plumes below melting Arctic sea ice

    Science.gov (United States)

    Peterson, Algot K.

    2018-02-01

    In sea ice, interconnected pockets and channels of brine are surrounded by fresh ice. Over time, brine is lost by gravity drainage and flushing. The timing of salt release and its interaction with the underlying water can impact subsequent sea ice melt. Turbulence measurements 1 m below melting sea ice north of Svalbard reveal anticorrelated heat and salt fluxes. From the observations, 131 salty plumes descending from the warm sea ice are identified, confirming previous observations from a Svalbard fjord. The plumes are likely triggered by oceanic heat through bottom melt. Calculated over a composite plume, oceanic heat and salt fluxes during the plumes account for 6 and 9 % of the total fluxes, respectively, while only lasting in total 0.5 % of the time. The observed salt flux accumulates to 7.6 kg m-2, indicating nearly full desalination of the ice. Bulk salinity reduction between two nearby ice cores agrees with accumulated salt fluxes to within a factor of 2. The increasing fraction of younger, more saline ice in the Arctic suggests an increase in desalination processes with the transition to the new Arctic.

  7. Melt extraction during heating and cooling of felsic crystal mushes in shallow volcanic systems: An experimental study

    Science.gov (United States)

    Pistone, M.; Baumgartner, L. P.; Sisson, T. W.; Bloch, E. M.

    2017-12-01

    The dynamics and kinetics of melt extraction in near-solidus, rheologically stalled, felsic crystal mushes (> 50 vol.% crystals) are essential to feeding many volcanic eruptions. At shallow depths (volatile-saturated and may be thermally stable for long time periods (104-107 years). In absence of deformation, residual melt can segregate from the mush's crystalline framework stimulated by: 1) gas injecting from hot mafic magmas into felsic mushes (heating / partial melting scenario), and 2) gas exsolving from the crystallizing mush (cooling / crystallizing scenario). The conditions and efficiency of melt extraction from a mush in the two scenarios are not well understood. Thus, we conducted high-temperature (700 to 850 °C) and -pressure (1.1 kbar) cold seal experiments (8-day duration) on synthetic felsic mushes, composed of water-saturated (4.2 wt.%) rhyodacite melt bearing different proportions of added quartz crystals (60, 70, and 80 vol%; 68 mm average particle size). High-spatial resolution X-ray tomography of run products show: 1) in the heating scenario (> 750 °C) melt has not segregated due to coalescence of vesicles (≤ 23 vol%) and large melt connectivity (> 7 vol% glass) / low pressure gradient for melt movement up to 80 vol% crystals; 2) in the cooling scenario (≤ 750 °C) vesicle (< 11 vol%) coalescence is limited or absent and limited amount of melt (3 to 11 vol%) segregated from sample center to its outer periphery (30 to 100 mm melt-rich lenses), testifying to the efficiency of melt extraction dictated by increasing crystallinity. These results suggest that silicic melt hosted within a crystal-rich mush can accumulate rapidly due to the buildup of modest gas pressures during crystallization at temperatures near the solidus.

  8. Electron beam melting of bearing materials

    Energy Technology Data Exchange (ETDEWEB)

    Goldschmied, G.; Schuler, A. (Technische Univ., Vienna (Austria). Inst. fuer Allgemeine Elektrotechnik); Elsinger, G.; Koroschetz, F. (MIBA Gleitlager AG, Laakirchen (Austria)); Tschegg, E.K. (Technische Univ., Vienna (Austria). Inst. fuer Angewandte und Technische Physik)

    1990-06-01

    This paper reports on a surface treatment method for the bearing materials AlSn6 which permits the use of this material without the overlay usually required. Microstructural refinement is achieved by means of a surface melting technique using an electron beam with successive rapid solidification. Extremely fine tin precipitates are formed in the melted surface layer which lead to significantly better tribological properties of the bearing material. Tests compared the tribological properties for AlSn6 bearings treated by the surface melting technique with those of untreated bearings. Whereas all untreated bearings failed by seizure after only 2 h of testing, 30% of the tested bearings which had been surface melted survived the entire testing program without damage.

  9. Extraction of scandium by organic substance melts

    International Nuclear Information System (INIS)

    Gladyshev, V.P.; Lobanov, F.I.; Zebreva, A.I.; Andreeva, N.N.; Manuilova, O.A.; Il'yukevich, Yu.A.

    1984-01-01

    Regularities of scandium extraction by the melts of octadecanicoic acid, n-carbonic acids of C 17 -C 20 commerical fraction and mixtures of tributylphosphate (TBP) with paraffin at (70+-1) deg C have been studied. The optimum conditions for scandium extraction in the melt of organic substances are determined. A scheme of the extraction by the melts of higher carbonic acids at ninitial metal concentrations of 10 -5 to 10 -3 mol/l has been suggested. The scandium compound has been isolated in solid form, its composition having been determined. The main advantages of extraction by melts are as follows: a possibility to attain high distribution coefficients, distinct separation of phases after extraction, the absence of emulsions, elimination of employing inflammable and toxic solvents, a possibility of rapid X-ray fluorescence determinatinon of scandium directly in solid extract

  10. Vertical melting of a stack of membranes

    Science.gov (United States)

    Borelli, M. E. S.; Kleinert, H.; Schakel, A. M. J.

    2001-02-01

    A stack of tensionless membranes with nonlinear curvature energy and vertical harmonic interaction is studied. At low temperatures, the system forms a lamellar phase. At a critical temperature, the stack disorders vertically in a melting-like transition.

  11. Selective Laser Ablation and Melting, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — In this project Advratech will develop a new additive manufacturing (AM) process called Selective Laser Ablation and Melting (SLAM). The key innovation in this...

  12. Energy Saving Melting and Revert Reduction Technology (E-SMARRT): Melting Efficiency Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Principal Investigator Kent Peaslee; Co-PI’s: Von Richards, Jeffrey Smith

    2012-07-31

    Steel foundries melt recycled scrap in electric furnaces and typically consume 35-100% excess energy from the theoretical energy requirement required to pour metal castings. This excess melting energy is multiplied by yield losses during casting and finishing operations resulting in the embodied energy in a cast product typically being three to six times the theoretical energy requirement. The purpose of this research project was to study steel foundry melting operations to understand energy use and requirements for casting operations, define variations in energy consumption, determine technologies and practices that are successful in reducing melting energy and develop new melting techniques and tools to improve the energy efficiency of melting in steel foundry operations.

  13. Theoretical work on melt-coolant interactions (steam explosions)

    International Nuclear Information System (INIS)

    Arnecke, G.; Jacobs, H.; Stehle, B.; Thurnay, K.; Vaeth, L.; Lummer, M.

    1995-01-01

    The code IVA3 is used for modelling the physical processes related to steam explosions, i.e. the premixing phase preceding the explosion as well as the explosion itself. This code has been replaced by the updated version IVA-KA in May 1994, which encompasses all model and code improvements performed till the beginning of 1994. The following further work on and with IVA-KA has been performed: 1. Inclusion of friction at inner and outer walls, improvement on the drag model, improvement of boundary conditions for outgoing flow, optional inclusion of improved water material data, improvement of the numerical procedure, correction of coding errors. 2. Three FARO-experiments (investigating the behaviour of molten material falling into water) were recalculated with IVA-KA. The time dependent pressure increase is reproduced very well for one experiment, but is not quite satisfactory for a second one. The third one cannot be simulated satisfactorily because of the presence of metallic zirconium in the melt, which is not being modelled by IVA-KA at present. 3. One PREMIX-experiment (similar to FARO, but at 1 bar ambient pressure and with smaller amounts of melt) is also being analyzed with IVA-KA. First results show a good representation of the material distribution during the penetration of the melt into the water. 4. One of the first two QUEOS-experiments performed at KfK has been simulated with IVA-KA. Some results are well reproduced by IVA-KA, but there may be a deficiency of the drag laws. (orig./HP)

  14. Basal melting driven by turbulent thermal convection

    Science.gov (United States)

    Rabbanipour Esfahani, Babak; Hirata, Silvia C.; Berti, Stefano; Calzavarini, Enrico

    2018-05-01

    Melting and, conversely, solidification processes in the presence of convection are key to many geophysical problems. An essential question related to these phenomena concerns the estimation of the (time-evolving) melting rate, which is tightly connected to the turbulent convective dynamics in the bulk of the melt fluid and the heat transfer at the liquid-solid interface. In this work, we consider a convective-melting model, constructed as a generalization of the Rayleigh-Bénard system, accounting for the basal melting of a solid. As the change of phase proceeds, a fluid layer grows at the heated bottom of the system and eventually reaches a turbulent convection state. By means of extensive lattice-Boltzmann numerical simulations employing an enthalpy formulation of the governing equations, we explore the model dynamics in two- and three-dimensional configurations. The focus of the analysis is on the scaling of global quantities like the heat flux and the kinetic energy with the Rayleigh number, as well as on the interface morphology and the effects of space dimensionality. Independently of dimensionality, we find that the convective-melting system behavior shares strong resemblances with that of the Rayleigh-Bénard one, and that the heat flux is only weakly enhanced with respect to that case. Such similarities are understood, at least to some extent, considering the resulting slow motion of the melting front (with respect to the turbulent fluid velocity fluctuations) and its generally little roughness (compared to the height of the fluid layer). Varying the Stefan number, accounting for the thermodynamical properties of the material, also seems to have only a mild effect, which implies the possibility of extrapolating results in numerically delicate low-Stefan setups from more convenient high-Stefan ones. Finally, we discuss the implications of our findings for the geophysically relevant problem of modeling Arctic ice melt ponds.

  15. Uniaxial Elongational viscosity of bidisperse polystyrene melts

    DEFF Research Database (Denmark)

    Nielsen, Jens Kromann; Rasmussen, Henrik K.; Hassager, Ole

    2006-01-01

    The startup and steady uniaxial elongational viscosity have been measured for three bidisperse polystyrene (PS) melts, consisting of blends of monodisperse PS with molecular weights of 52 kg/mole or 103 kg/mole and 390 kg/mole. The bidisperse melts have a maximum in the steady elongational...... viscosity, of up to a factor of 7 times the Trouton limit of 3 times the zero-shear viscosity....

  16. Shock induced melting of lead (experimental study)

    International Nuclear Information System (INIS)

    Mabire, Catherine; Hereil, Pierre L.

    2002-01-01

    To investigate melting on release of lead, two shock compression measurements have been carried out at 51 GPa. In the first one, a pyrometric measurement has been performed at the Pb/LiF interface. In the second one, the Pb/LiF interface velocity has been recorded using VISAR measurement technique. VISAR and radiance profile are in good agreement and seem to show melting on release of lead

  17. Vacancies in quantal Wigner crystals near melting

    International Nuclear Information System (INIS)

    Barraza, N.; Colletti, L.; Tosi, M.P.

    1999-04-01

    We estimate the formation energy of lattice vacancies in quantal Wigner crystals of charged particles near their melting point at zero temperature, in terms of the crystalline Lindemann parameter and of the static dielectric function of the fluid phase near freezing. For both 3D and 2D crystals of electrons our results suggest the presence of vacancies in the ground state at the melting density. (author)

  18. Electrodepositions on Tantalum in Alkali Halide Melts

    DEFF Research Database (Denmark)

    Barner, Jens H. Von; Jensen, Annemette Hindhede; Christensen, Erik

    2013-01-01

    Surface layers of tantalum metal were electrodeposited on steel from K2TaF7-LiF-NaF-KF melts. With careful control of the oxide contents dense and adherent deposits could be obtained by pulse plating. In NaCl-KCl-NaF-Na2CO3 and NaCl-KCl-Na2CO3 melts carbonate ions seems to be reduced to carbon in...

  19. Electrodepositions on Tantalum in alkali halide melts

    DEFF Research Database (Denmark)

    Barner, Jens H. Von; Jensen, Annemette Hindhede; Christensen, Erik

    2012-01-01

    Surface layers of tantalum metal were electrodeposited on steel from K 2TaF7-LiF-NaF-KF melts. With careful control of the oxide contents dense and adherent deposits could be obtained by pulse plating. In NaCl-KCl-NaF-Na2CO3 and NaCl-KCl-Na2CO 3 melts carbonate ions seems to be reduced to carbon ...

  20. Depth and degree of melting of komatiites

    Science.gov (United States)

    Herzberg, Claude

    1992-04-01

    High pressure melting experiments have permitted new constraints to be placed on the depth and degree of partial melting of komatiites. Komatiites from Gorgona Island were formed by relatively low degrees of pseudoinvariant melting involving L + Ol + Opx + Cpx + Gt on the solidus at 40 kbar, about 130 km depth. Munro-type komatiites were separated from a harzburgite residue (L + Ol + Opx) at pressures that were poorly constrained, but were probably around 50 kbar, about 165 km depth; the degree of partial melting was less than 40 percent. Secular variations in the geochemistry of komatiites could have formed in response to a reduction in the temperature and pressure of melting with time. The 3.5 Ga Barberton komatiites and the 2.7 Ga Munro-type komatiities could have formed in plumes that were hotter than the present-day mantle by 500 deg and 300 deg, respectively. When excess temperatures are this size, melting is deeper and volcanism changes from basaltic to momatiitic. The komatiities from Gorgona Island, which are Mesozoic in age, may be representative of komatiities that are predicted to occur in oceanic plateaus of Cretaceous age throughout the Pacific (Storey et al., 1991).

  1. The melting and solidification of nanowires

    International Nuclear Information System (INIS)

    Florio, B. J.; Myers, T. G.

    2016-01-01

    A mathematical model is developed to describe the melting of nanowires. The first section of the paper deals with a standard theoretical situation, where the wire melts due to a fixed boundary temperature. This analysis allows us to compare with existing results for the phase change of nanospheres. The equivalent solidification problem is also examined. This shows that solidification is a faster process than melting; this is because the energy transfer occurs primarily through the solid rather than the liquid which is a poorer conductor of heat. This effect competes with the energy required to create new solid surface which acts to slow down the process, but overall conduction dominates. In the second section, we consider a more physically realistic boundary condition, where the phase change occurs due to a heat flux from surrounding material. This removes the singularity in initial melt velocity predicted in previous models of nanoparticle melting. It is shown that even with the highest possible flux the melting time is significantly slower than with a fixed boundary temperature condition.

  2. The melting and solidification of nanowires

    Science.gov (United States)

    Florio, B. J.; Myers, T. G.

    2016-06-01

    A mathematical model is developed to describe the melting of nanowires. The first section of the paper deals with a standard theoretical situation, where the wire melts due to a fixed boundary temperature. This analysis allows us to compare with existing results for the phase change of nanospheres. The equivalent solidification problem is also examined. This shows that solidification is a faster process than melting; this is because the energy transfer occurs primarily through the solid rather than the liquid which is a poorer conductor of heat. This effect competes with the energy required to create new solid surface which acts to slow down the process, but overall conduction dominates. In the second section, we consider a more physically realistic boundary condition, where the phase change occurs due to a heat flux from surrounding material. This removes the singularity in initial melt velocity predicted in previous models of nanoparticle melting. It is shown that even with the highest possible flux the melting time is significantly slower than with a fixed boundary temperature condition.

  3. The melting and solidification of nanowires

    Energy Technology Data Exchange (ETDEWEB)

    Florio, B. J., E-mail: brendan.florio@ul.ie [University of Limerick, Mathematics Applications Consortium for Science and Industry (MACSI), Department of Mathematics and Statistics (Ireland); Myers, T. G., E-mail: tmyers@crm.cat [Centre de Recerca Matemàtica (Spain)

    2016-06-15

    A mathematical model is developed to describe the melting of nanowires. The first section of the paper deals with a standard theoretical situation, where the wire melts due to a fixed boundary temperature. This analysis allows us to compare with existing results for the phase change of nanospheres. The equivalent solidification problem is also examined. This shows that solidification is a faster process than melting; this is because the energy transfer occurs primarily through the solid rather than the liquid which is a poorer conductor of heat. This effect competes with the energy required to create new solid surface which acts to slow down the process, but overall conduction dominates. In the second section, we consider a more physically realistic boundary condition, where the phase change occurs due to a heat flux from surrounding material. This removes the singularity in initial melt velocity predicted in previous models of nanoparticle melting. It is shown that even with the highest possible flux the melting time is significantly slower than with a fixed boundary temperature condition.

  4. On melting criteria for complex plasma

    International Nuclear Information System (INIS)

    Klumov, Boris A

    2011-01-01

    The present paper considers melting criteria for a plasma crystal discovered in dust plasma in 1994. Separate discussions are devoted to three-dimensional (3D) and two-dimensional (2D) systems. In the 3D case, melting criteria are derived based on the properties of local order in a system of microparticles. The order parameters are constructed from the cumulative distributions of the microparticle probability distributions as functions of various rotational invariants. The melting criteria proposed are constructed using static information on microparticle positions: a few snapshots of the system that allow for the determination of particle coordinates are enough to determine the phase state of the system. It is shown that criteria obtained in this way describe well the melting and premelting of 3D complex plasmas. In 2D systems, a system of microparticles interacting via a screened Coulomb (i.e., Debye-Hueckel or Yukawa) potential is considered as an example, using molecular dynamics simulations. A number of new order parameters characterizing the melting of 2D complex plasmas are proposed. The order parameters and melting criteria proposed for 2D and 3D complex plasmas can be applied to other systems as well. (methodological notes)

  5. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, A., E-mail: alan.peacock@ipp.mpg.de [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Girlinger, A. [MAN Diesel and Turbo SE D-94469 Deggendorf (Germany); Vorkoeper, A. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 17491 Greifswald (Germany); Boscary, J.; Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Hurd, F. [European Commission c/o Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany); Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G. [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, 85748 Garching (Germany)

    2011-10-15

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m{sup 2}, are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m{sup 2} and a maximum heat load of 200 kW/m{sup 2} with a water velocity of approximately 2 m s{sup -1}. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m{sup 2} for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  6. The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

    International Nuclear Information System (INIS)

    Peacock, A.; Girlinger, A.; Vorkoeper, A.; Boscary, J.; Greuner, H.; Hurd, F.; Mendelevitch, B.; Pirsch, H.; Stadler, R.; Zangl, G.

    2011-01-01

    320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m 2 , are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel and Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m 2 and a maximum heat load of 200 kW/m 2 with a water velocity of approximately 2 m s -1 . High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m 2 for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.

  7. Rapakivi texture formation via disequilibrium melting in a contact partial melt zone, Antarctica

    Science.gov (United States)

    Currier, R. M.

    2017-12-01

    In the McMurdo Dry Valleys of Antarctica, a Jurassic aged dolerite sill induced partial melting of granite in the shallow crust. The melt zone can be traced in full, from high degrees of melting (>60%) along the dolerite contact, to no apparent signs of melting, 10s of meters above the contact. Within this melt zone, the well-known rapakivi texture is found, arrested in various stages of development. High above the contact, and at low degrees of melting, K-feldspar crystals are slightly rounded and unmantled. In the lower half of the melt zone, mantles of cellular textured plagioclase appear on K-feldspar, and thicken towards the contact heat source. At the highest degrees of melting, cellular-textured plagioclase completely replaces restitic K-feldspar. Because of the complete exposure and intact context, the leading models of rapakivi texture formation can be tested against this system. The previously proposed mechanisms of subisothermal decompression, magma-mixing, and hydrothermal exsolution all fail to adequately describe rapakivi generation in this melt zone. Preferred here is a closed system model that invokes the production of a heterogeneous, disequilibrium melt through rapid heating, followed by calcium and sodium rich melt reacting in a peritectic fashion with restitic K-feldspar crystals. This peritectic reaction results in the production of plagioclase of andesine-oligoclase composition—which is consistent with not just mantles in the melt zone, but globally as well. The thickness of the mantle is diffusion limited, and thus a measure of the diffusive length scale of sodium and calcium over the time scale of melting. Thermal modeling provides a time scale of melting that is consistent with the thickness of observed mantles. Lastly, the distribution of mantled feldspars is highly ordered in this melt zone, but if it were mobilized and homogenized—mixing together cellular plagioclase, mantled feldspars, and unmantled feldspars—the result would be

  8. Water

    CSIR Research Space (South Africa)

    Van Wyk, Llewellyn V

    2010-08-01

    Full Text Available Water scarcity is without a doubt on of the greatest threats to the human species and has all the potential to destabilise world peace. Falling water tables are a new phenomenon. Up until the development of steam and electric motors, deep groudwater...

  9. Water

    OpenAIRE

    Hertie School of Governance

    2010-01-01

    All human life depends on water and air. The sustainable management of both is a major challenge for today's public policy makers. This issue of Schlossplatz³ taps the streams and flows of the current debate on the right water governance.

  10. Glacier Melt Detection in Complex Terrain Using New AMSR-E Calibrated Enhanced Daily EASE-Grid 2.0 Brightness Temperature (CETB) Earth System Data Record

    Science.gov (United States)

    Ramage, J. M.; Brodzik, M. J.; Hardman, M.

    2016-12-01

    Passive microwave (PM) 18 GHz and 36 GHz horizontally- and vertically-polarized brightness temperatures (Tb) channels from the Advanced Microwave Scanning Radiometer for EOS (AMSR-E) have been important sources of information about snow melt status in glacial environments, particularly at high latitudes. PM data are sensitive to the changes in near-surface liquid water that accompany melt onset, melt intensification, and refreezing. Overpasses are frequent enough that in most areas multiple (2-8) observations per day are possible, yielding the potential for determining the dynamic state of the snow pack during transition seasons. AMSR-E Tb data have been used effectively to determine melt onset and melt intensification using daily Tb and diurnal amplitude variation (DAV) thresholds. Due to mixed pixels in historically coarse spatial resolution Tb data, melt analysis has been impractical in ice-marginal zones where pixels may be only fractionally snow/ice covered, and in areas where the glacier is near large bodies of water: even small regions of open water in a pixel severely impact the microwave signal. We use the new enhanced-resolution Calibrated Passive Microwave Daily EASE-Grid 2.0 Brightness Temperature (CETB) Earth System Data Record product's twice daily obserations to test and update existing snow melt algorithms by determining appropriate melt thresholds for both Tb and DAV for the CETB 18 and 36 GHz channels. We use the enhanced resolution data to evaluate melt characteristics along glacier margins and melt transition zones during the melt seasons in locations spanning a wide range of melt scenarios, including the Patagonian Andes, the Alaskan Coast Range, and the Russian High Arctic icecaps. We quantify how improvement of spatial resolution from the original 12.5 - 25 km-scale pixels to the enhanced resolution of 3.125 - 6.25 km improves the ability to evaluate melt timing across boundaries and transition zones in diverse glacial environments.

  11. Microwave melt and offgas analysis results from a Ferro Corporation reg-sign glass frit

    International Nuclear Information System (INIS)

    Phillips, J.A.; Hoffman, C.R.; Knutson, P.T.

    1995-03-01

    In support of the Residue Treatment Technology (RTT) Microwave Solidification project, Waste Projects and Surface Water personnel conducted a series of experiments to determine the feasibility of encapsulating a surrogate sludge waste using the microwave melter. The surrogate waste was prepared by RTT and melted with five varying compositions of low melting glass frit supplied by the Ferro Corporation. Samples were melted using a 50% waste/50% glass frit and a 47.5% waste/47.5% glass frit/5% carbon powder. This was done to evaluate the effectiveness of carbon at reducing a sulfate-based surface scale which has been observed in previous experiments and in full-scale testing. These vitrified samples were subsequently submitted to Environmental Technology for toxicity characteristic leaching procedure (TCLP) testing. Two of the five frits tested in this experiment merit further evaluation as raw materials for the microwave melter. Ferro frit 3110 with and without carbon powder produced a crystalline product which passed TCLP testing. The quality of the melt product could be improved by increasing the melting temperature from 900 degrees C to approximately 1150-1200 degrees C. Ferro frit 3249 produced the optimal quality of glass based on visual observations, but failed TCLP testing for silver when melted without carbon powder. This frit requires a slightly higher melting temperature (≥ 1200 degrees C) compared to frit 3110 and produces a superior product. In conjunction with this work, Surface Water personnel conducted offgas analyses using a Thermal Desorption Mass Spectrometer (TDMS) on selected formulations. The offgas analyses identified and quantified water vapor (H 2 O), oxygen (O 2 ) and carbon oxides (CO and CO 2 ), sulfur (S) and sulfur oxides (SO and SO 2 ), and nitrogen (N 2 ) and nitrogen oxides (NO and NO 2 ) that volatilized during glass formation

  12. Silicate melts density, buoyancy relations and the dynamics of magmatic processes in the upper mantle

    Science.gov (United States)

    Sanchez-Valle, Carmen; Malfait, Wim J.

    2016-04-01

    Although silicate melts comprise only a minor volume fraction of the present day Earth, they play a critical role on the Earth's geochemical and geodynamical evolution. Their physical properties, namely the density, are a key control on many magmatic processes, including magma chamber dynamics and volcanic eruptions, melt extraction from residual rocks during partial melting, as well as crystal settling and melt migration. However, the quantitative modeling of these processes has been long limited by the scarcity of data on the density and compressibility of volatile-bearing silicate melts at relevant pressure and temperature conditions. In the last decade, new experimental designs namely combining large volume presses and synchrotron-based techniques have opened the possibility for determining in situ the density of a wide range of dry and volatile-bearing (H2O and CO2) silicate melt compositions at high pressure-high temperature conditions. In this contribution we will illustrate some of these progresses with focus on recent results on the density of dry and hydrous felsic and intermediate melt compositions (rhyolite, phonolite and andesite melts) at crustal and upper mantle conditions (up to 4 GPa and 2000 K). The new data on felsic-intermediate melts has been combined with in situ data on (ultra)mafic systems and ambient pressure dilatometry and sound velocity data to calibrate a continuous, predictive density model for hydrous and CO2-bearing silicate melts with applications to magmatic processes down to the conditions of the mantle transition zone (up to 2773 K and 22 GPa). The calibration dataset consist of more than 370 density measurements on high-pressure and/or water-and CO2-bearing melts and it is formulated in terms of the partial molar properties of the oxide components. The model predicts the density of volatile-bearing liquids to within 42 kg/m3 in the calibration interval and the model extrapolations up to 3000 K and 100 GPa are in good agreement

  13. Alignment of in-vessel components by metrology defined adaptive machining

    International Nuclear Information System (INIS)

    Wilson, David; Bernard, Nathanaël; Mariani, Antony

    2015-01-01

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

  14. The control and data acquisition system of a laser in-vessel viewing system

    International Nuclear Information System (INIS)

    Pereira, Rita C.; Cruz, Nuno; Neri, C.; Riva, M.; Correia, C.; Varandas, C.A.F.

    2000-01-01

    This paper presents the dedicated control and data acquisition system (CADAS) of a new laser in-vessel viewing system that has been developed for inspection purposes in fusion experiments. CADAS is based on a MC68060 microprocessor and on-site developed VME instrumentation. Its main aims are to simultaneously control the laser alignment system as well as the laser beam deflection for in-vessel scanning, acquire a high-resolution image and support real-time data flow rates up to 2 Mbyte/s from the acquisition modules to the hard disk and network. The hardware (modules for control and alignment acquisition, scanning acquisition and monitoring) as well as the three levels of software are described

  15. Alignment of in-vessel components by metrology defined adaptive machining

    Energy Technology Data Exchange (ETDEWEB)

    Wil