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

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.

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.

Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

International Nuclear Information System (INIS)

Frid, W.

1988-04-01

Discharge of the molten core debris from a pressurized reactor vessel has been recognized as an important accident scenario for pressurized water reactors. Recent high-pressure melt streaming experiments conducted at Sandia National Laboratories, designed to study cavity and containment events related to melt ejection, have resulted in two important observations: (1) Expansion and breakup of the ejected molten jet. (2) Significant aerosol generation during the ejection process. The expansion and breakup of the jet in the experiments are attributed to rapid evolution of the pressurizing gas (nitrogen or hydrogen) dissolved in the melt. It has been concluded that aerosol particles may be formed by condensation of melt vapor and mechanical breakup of the melt and generation. It was also shown that the above stated phenomena are likely to occur in reactor accidents. This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals, and gas bubble nucleation in molten metals are relevant problems that are addressed in this work

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

Ex-Vessel Steam Explosion Analysis of Central Melt Pour Scenario

International Nuclear Information System (INIS)

Ursic, M.; Leskovar, M.

2008-01-01

An ex-vessel steam explosion may develop during a severe reactor accident when the reactor vessel fails and the molten core interacts with the coolant in the reactor cavity. At this process part of the corium energy is intensively transferred to water in a very short time scale. The water vaporizes at high pressure and expands, doing work on its surrounding. Although the steam explosion has probably a low probability of occurrence, it is an important nuclear safety issue in case of a severe reactor accident. Namely, the formed very high pressure region induces dynamic loadings on the surrounding structures that may potentially lead to an early release of the radioactive material into the environment. Although the steam explosion events have being studied for several years, the level of the process and consequences understanding is still not adequate. To increase the level of confidence the OECD programme SERENA (Steam Explosion REsolution for Nuclear Applications) was established in 2002. The objectives of the program were to evaluate capabilities of the current generation of the FCI (Fuel-Coolant Interaction) computer codes in predicting the steam explosion induced loads, identifying key FCI phenomena and associated uncertainties impacting the predictability of the steam explosion energetics in the reactor situations and proposing confirmatory research to reduce the uncertainties to acceptable levels for the steam explosion risk assessment. To get a better insight into the most challenging ex-vessel steam explosions, analyses for different locations of the melt release, the cavity water sub-cooling, the primary system pressure overpressure and the triggering time were preformed for a typical pressurized water reactor cavity. The results of some scenarios revealed that significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. Among the performed analyses for the central melt pour scenarios, the maximum pressure loads were

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-03-01

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

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

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)

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.

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.)

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

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)

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

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.

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)

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).

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

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

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

Final report for the 'Melt-Vessel Interactions' Project. European Union R and TD Program 4th Framework. MVI project final research report

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Green, J.; Kolb, G.; Karbojian, A.; Theerthan, S.A.; Gubaidulline, A.; Bonnet, J.M.; Rouge, S.; Narcoux, M.; Liegeois, A.; Turland, B.D.; Dobson, G.P.; Siccama, A.; Ikonen, K.; Parozzi, F.; Kolev, N.; Caira, M.

1999-04-01

The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the Final Report for the Project, in which a summary of the progress achieved in the experimental program is provided. We have, however, included some aspects of the modeling activities. Volume 2 of the Final report describes the progress achieved in the modeling program. The progress achieved in the experimental and modeling parts of the Project has led to the resolution of some of the issues of melt vessel interaction. Considerable progress was also achieved towards resolution of the remaining issues

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

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

International Nuclear Information System (INIS)

Corradini, Michael L.

2006-01-01

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

Prediction of thermoplastic failure of a reactor pressure vessel under a postulated core melt accident

International Nuclear Information System (INIS)

Duijvestijn, G.; Birchley, J.; Reichlin, K.

1997-01-01

This paper presents the lower head failure calculations performed for a postulated accident scenario in a commercial nuclear power plant. A postulated one inch break in the primary coolant circuit leads to dryout and subsequent meltdown of the core. The reference plant is a pressurized water reactor without penetrations in the reactor vessel lower head. The molten core material accumulates in the lower head, eventually causing failure of the vessel. The analysis investigates flow conditions in the melt pool, temperature evolution in the reactor vessel wall, and structure mechanical evaluation of the vessel under strong thermal loads and a range of internal pressures. The calculations were performed using the ADINA finite element codes. The analysis focusses on the failure processes, time and mode of failure. The most likely mode of failure at low pressure is global rupture due to gradual accumulation of creep strain over a large part of the heated area. In contrast, thermoplasticity becomes important at high pressure or following a pressure spike and can lead to earlier local failure. In situations in which part of the heat load is concentrated over a small area, resulting in a hot spot, local failure occurs, but not until the temperatures are close to the melting point. At low pressure, in particular, the hot spot area remains intact until the structure is molten across more than half of the thickness. (author) 14 figs., 16 refs

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

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

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

Final report for the 'Melt-Vessel Interactions' Project. European Union R and TD Program 4th Framework. MVI project final research report

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Green, J.; Kolb, G.; Karbojian, A.; Theerthan, S.A.; Gubaidulline, A. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety; Helle, M.; Kymaelaeinen, O.; Tuomisto, H. [IVO Power Engineering Ltd., Vantaa (Finland); Bonnet, J.M.; Rouge, S.; Narcoux, M.; Liegeois, A. [CEA - Grenoble (France); Turland, B.D.; Dobson, G.P. [AEA Technology plc, Dorchester (United Kingdom); Siccama, A. [ECN Nuclear Research, Petten (Netherlands); Ikonen, K. [VTT Energy, Helsinki (Finland); Parozzi, F. [ENEL - SRI/PAM/GRA, Segrate, MI (Italy); Kolev, N. [Siemens AG, Erlangen (Germany); Caira, M. [Univ. of Roma (Italy)

1999-04-01

The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the Final Report for the Project, in which a summary of the progress achieved in the experimental program is provided. We have, however, included some aspects of the modeling activities. Volume 2 of the Final report describes the progress achieved in the modeling program. The progress achieved in the experimental and modeling parts of the Project has led to the resolution of some of the issues of melt vessel interaction. Considerable progress was also achieved towards resolution of the remaining issues.

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

International Nuclear Information System (INIS)

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

2008-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-15

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

Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

International Nuclear Information System (INIS)

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

1995-07-01

Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs

Quantification of the ex-vessel severe accident risks for the Swedish boiling water reactors. A scoping study performed for the APRI project

Energy Technology Data Exchange (ETDEWEB)

Okkonen, T; Dinh, T N; Bui, V A; Sehgal, B R [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Energy Systems Technology

1995-07-01

Results of a scoping study to quantify the ex-vessel severe accident risks for the Swedish BWRs are reported. The study considers that a pool of water is established in the containment prior to vessel failure, as prescribed by the accident management scheme for the newer Swedish BWRs. The integrated methodology developed and employed combines probabilistic and deterministic treatment of the various melt-structure-water interaction processes occurring in sequence. The potential steam explosion, and the melt attack on the containment basemat, are treated with enveloping analyses. Uncertain parameters in the models and the initial conditions are treated with Monte Carlo simulations. Independent models are developed for melt coolability and possible attack on the concrete basemat. It is found that, with current models, the melt discharge scenarios, in which a large amount of accumulated melt may be released from the vessel, could subject the containment to large steam explosion loads. However, the uncertainties are so large that no definite conclusion can be drawn. The assessment of ex-vessel core debris coolability is disturbed by similar phenomenological uncertainties. Presently, coolability of the core debris can not be demonstrated. 133 refs.

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

Standard Guide for Use of Melt Wire Temperature Monitors for Reactor Vessel Surveillance, E 706 (IIIE)

CERN Document Server

American Society for Testing and Materials. Philadelphia

2006-01-01

1.1 This guide describes the application of melt wire temperature monitors and their use for reactor vessel surveillance of light-water power reactors as called for in Practice E 185. 1.2 The purpose of this guide is to recommend the selection and use of the common melt wire technique where the correspondence between melting temperature and composition of different alloys is used as a passive temperature monitor. Guidelines are provided for the selection and calibration of monitor materials; design, fabrication, and assembly of monitor and container; post-irradiation examinations; interpretation of the results; and estimation of uncertainties. 1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use. (See Note 1.)

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)

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

An experimental study on feasibility of ex-vessel cooling through the external guide vessel

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Kim, Jong-Hwan; Park, Rae-Jun; Kim, Sang-Baik

2000-01-01

This paper presents the results of a series of experiments for assessing the efficacy of ex-vessel cooling through the external guide vessel during a severe accident. Four tests were performed in the LAVA test facility at KAERI, varying the boundary conditions at the outer surface of the vessel. The first test was a dry condition test conducted without cooling the outside of the vessel. On the other hand, in the second test, the cooling of the vessel surface was produced by gravity-driven forced injection of water along the annular gap of 25 mm between the vessel and the external guide vessel. Water flow rate was about 0.85 kg/s and total mass of available water was 300 kg. For the evaluation of the water flow rate effect, the third test was performed with a pool type cooling in the annulus without any circulation of water. These two external cooling tests were performed under elevated pressure of about 1.6 MPa. Finally, the fourth test was conducted under atmospheric pressure to evaluate the effect of system pressure on boiling heat transfer characteristics. In the dry test and the pool type ex-vessel cooling test performed under atmospheric pressure, the vessel was failed by a melt penetration at about 40 degree upper position from the vessel bottom, which is coincident with the boundary of the Al 2 O 3 /Fe melt separated layers. On the other hand, in both of the ex-vessel cooling tests conducted under elevated pressure of about 1.6 MPa, the vessel didn't fail. Compared with the pool boiling test, the vessel experienced effective cooling due to the inlet flow in the forced flow test. Synthesized the results of the tests, it was shown that the heat removal with ex-vessel cooling through the guide vessel is feasible, but the additional evaluations should be performed to guarantee enough thermal margin. (author)

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

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)

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

International Nuclear Information System (INIS)

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

2005-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-10-01

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

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

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.

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.

Analysis of ex-vessel steam explosion with MC3D

International Nuclear Information System (INIS)

Leskovar, M.; Mavko, B.

2007-01-01

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. 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. In the paper, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which was developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed varying the location of the melt release (central, left and right side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. The performed analysis shows that for some ex-vessel steam explosion scenarios significantly higher pressure loads are predicted than obtained in the OECD programme SERENA Phase 1. (author)

Behavior of a corium jet in high pressure melt ejection from a reactor pressure vessel

International Nuclear Information System (INIS)

Frid, W.

1987-01-01

This report provides results from analytical and experimental investigations on the behavior of a gas supersaturated molten jet expelled from a pressurized vessel. Aero-hydrodynamic stability of liquid jets in gas, stream degassing of molten metals and gas bubble nucleation in molten metals are relevant problems which are addressed in this work. Models are developed for jet expansion, primary breakup of the jet and secondary fragmentation of melt droplets resulting from violent effervescence of dissolved gas. The jet expansion model is based on a general relation for bubble growth which includes both inertia-controlled and diffusion-controlled growth phases. The jet expansion model is able to predict the jet void fraction, jet radius as a function of axial distance from the pressure vessel, bubble size and bubble pressure. The number density of gas bubbles in the melt, which is a basic parameter in the model, was determined experimentally and is about 10 8 per m 3 of liquid. The primary breakup of the jet produces a spray of droplets, about 2-3 mm in diameter. Parametric calculations for a TMLB' reactor accident sequence show that the corium jet is disrupted within a few initial jet diameters from the reactor vessel and that the radius of corium spray at the level of the reactor cavity floor is in the range of 0.8 to 2.6 m. (orig./HP)

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

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.)

Method of detecting water leakage in radioactive waste containing vessel

International Nuclear Information System (INIS)

Ishioka, Hitoshi; Takao, Yoshiaki; Hayakawa, Kiyoshige.

1989-01-01

Lower level radioactive wastes formed upon operation of nuclear facilities are processed by underground storage. In this case, a plurality of drum cans packed with radioactive wastes are contained in a vessel and a water soluble dye material is placed at the inside of the vessel. The method of placing the water soluble dye material at the inside of the vessel includes a method of coating the material on the inner surface of the vessel and a method of mixing the material in sands to be filled between each of the drum cans. Then, leakage of water soluble dye material is detected when water intruding from the outside into the vessel is again leached out of the vessel, to detect the water leakage from the inside of the vessel. In this way, it is possible to find a water-invaded vessel before corrosion of the drum can by water intruded into the vessel and leakage of nuclides in the drum can. Accordingly, it is possible to apply treatment such as repair before occurrence of accident and can maintain the safety of radioactive water processing facilities. (I.S.)

Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

1994-10-01

As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

Introduction to the modified TROI test facility for fuel coolant interaction under a submerged reactor vessel

International Nuclear Information System (INIS)

Na, Young Su; Hong, Seong-Wan; Song, Jin Ho; Hong, Seong-Ho

2014-01-01

The molten Fuel-Coolant Interaction (FCI) can threaten the integrity of the reactor cavity under a severe accident. A steam explosion can be occurred by the rapid energy transfer in the high-temperature corium melt jet penetrating into water, which makes the dynamic load applying to the surrounding structure. Before a steam explosion, the corium melt jet breaks into small-sized particles, and the steam is generated continuously by the film boiling on the hot surface of the melt contacting with water. The premixing phase consisting of the corium melt, water, and steam can determine the intensity of the steam explosion. Unfortunately, the previous experimental studies on the FCI phenomena have carried out under a free fall of the corium melt jet in a gas phase before interacting with water. The previous TROI (Test for Real cOrium Interaction with water) test facility, that is a well-known test facility for the FCI phenomena in the world, has observed a steam explosion under a free fall of a corium melt jet in a gas phase before contacting a coolant since 2000, which is changing to simulate the FCI phenomena under a submerged reactor vessel. This study introduces the modified TROI test facility as shown in Fig. 1 and the considerations for the experiment with success. The previous TROI test facility, that has observed the molten Fuel-Coolant Interaction (FCI) with a free fall of the prototypic corium melt in a gas phase before contacting a coolant, was modified to simulate the FCI phenomena under a submerged reactor vessel for the assessment of the In-Vessel Retention (IVR) concept, i.e., without a free-fall distance of the corium melt before contacting water. The superheated prototypic corium melt created by the cold crucible melting method moves on a releasing valve newly installed just above the water level in the interaction vessel. The corium melt will stay on a releasing valve in less than 0.2 seconds to reduce heat loss for preventing the solidification, and

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

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.

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

Analysis of Reactor Vessel Lower Head Penetration Tube Failure

International Nuclear Information System (INIS)

Stempniewicz, Marek

1999-01-01

This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

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)

Melting and solidification characteristics of a mixture of two types of latent heat storage material in a vessel

Science.gov (United States)

Yu, JikSu; Horibe, Akihiko; Haruki, Naoto; Machida, Akito; Kato, Masashi

2016-11-01

In this study, we investigated the fundamental melting and solidification characteristics of mannitol, erythritol, and their mixture (70 % by mass mannitol: 30 % by mass erythritol) as potential phase-change materials (PCMs) for latent heat thermal energy storage systems, specifically those pertaining to industrial waste heat, having temperatures in the range of 100-250 °C. The melting point of erythritol and mannitol, the melting peak temperature of their mixture, and latent heat were measured using differential scanning calorimetry. The thermal performance of the mannitol mixture was determined during melting and solidification processes, using a heat storage vessel with a pipe heat exchanger. Our results indicated phase-change (fusion) temperatures of 160 °C for mannitol and 113 and 150 °C for the mannitol mixture. Nondimensional correlation equations of the average heat transfer during the solidification process, as well as the temperature and velocity efficiencies of flowing silicon oil in the pipe and the phase-change material (PCM), were derived using several nondimensional parameters.

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)

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)

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

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.

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

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.

Effects of air vessel on water hammer in high-head pumping station

International Nuclear Information System (INIS)

Wang, L; Wang, F J; Zou, Z C; Li, X N; Zhang, J C

2013-01-01

Effects of air vessel on water hammer process in a pumping station with high-head were analyzed by using the characteristics method. The results show that the air vessel volume is the key parameter that determines the protective effect on water hammer pressure. The maximum pressure in the system declines with increasing air vessel volume. For a fixed volume of air vessel, the shape of air vessel and mounting style, such as horizontal or vertical mounting, have little effect on the water hammer. In order to obtain good protection effects, the position of air vessel should be close to the outlet of the pump. Generally, once the volume of air vessel is guaranteed, the water hammer of a entire pipeline is effectively controlled

Effects of air vessel on water hammer in high-head pumping station

Science.gov (United States)

Wang, L.; Wang, F. J.; Zou, Z. C.; Li, X. N.; Zhang, J. C.

2013-12-01

Effects of air vessel on water hammer process in a pumping station with high-head were analyzed by using the characteristics method. The results show that the air vessel volume is the key parameter that determines the protective effect on water hammer pressure. The maximum pressure in the system declines with increasing air vessel volume. For a fixed volume of air vessel, the shape of air vessel and mounting style, such as horizontal or vertical mounting, have little effect on the water hammer. In order to obtain good protection effects, the position of air vessel should be close to the outlet of the pump. Generally, once the volume of air vessel is guaranteed, the water hammer of a entire pipeline is effectively controlled.

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2007-04-15

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

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

International Nuclear Information System (INIS)

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

2007-04-01

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

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)

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

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

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

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.

Assessment of reactor vessel integrity (ARVI)

Energy Technology Data Exchange (ETDEWEB)

Sehgal, B.R. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden)]. E-mail: sehgal@ne.kth.se; Karbojian, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Giri, A. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Kymaelaeinen, O. [FortumEngNP (Finland); Bonnet, J.M. [CEA (France); Ikkonen, K. [Division of Nuclear Power Safety (NPS), Royal Institute of Technology (KTH), Drottning Kristinas Vaeg 33A, 10044 Stockholm (Sweden); Sairanen, R. [VTT (Finland); Bhandari, S. [FRAMATOME (France); Buerger, M. [USTUTT (Germany); Dienstbier, J. [NRI Rez (Czech Republic); Techy, Z. [VEIKI (Hungary); Theofanous, T. [UCSB (United States)

2005-02-01

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.

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.

Cooling of pressurized water nuclear reactor vessels

International Nuclear Information System (INIS)

Curet, H.D.

1978-01-01

The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

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)

Ex-vessel corium coolability sensitivity study with the CORQUENCH code

International Nuclear Information System (INIS)

Robb, Kevin; Corradini, Michael

2009-01-01

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

Evaluation of an experiment modelling heat transfer from the melt pool for use in VVER 440/213 reactors

International Nuclear Information System (INIS)

Skop, J.

2003-12-01

The strategy of confining core melt within the reactor vessel is among promising strategies to mitigate severe accidents of VVER 440/213 reactors. This strategy consists in residual heat removal from the melt by external vessel cooling from the outside, using water from the flooded reactor downcomer. This approach can only be successful if the critical heat flux on the external vessel surface is not exceeded. This can be assessed based on the parameters of heat transfer from the core melt pool in the conditions of natural circulation within the pool. Those parameters are the subject of the report. A basic description of the terms and physical basis of the strategy of confining core melt inside the vessel is given in Chapter 2, which also briefly explains similarity theory, based on which the results obtained on experimental facilities, using simulation materials, can be related to the actual situation inside a real reactor. Chapter 3 presents an overview of experimental work addressing the characteristics of heat transfer from the core melt pool in natural circulation conditions and a description of the experimental facilities. An overview of the results emerging from the experiments and their evaluation with respect to their applicability to reactors in Czech nuclear power plants are given in Chapter 4

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

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

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

1991-01-01

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

33 CFR 64.16 - Duration of marking on sunken vessels in navigable waters.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Duration of marking on sunken vessels in navigable waters. 64.16 Section 64.16 Navigation and Navigable Waters COAST GUARD, DEPARTMENT... Sunken Vessels and Other Obstructions § 64.16 Duration of marking on sunken vessels in navigable waters...

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

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

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.

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).

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

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.

Development of LILAC-meltpool for the thermo-hydraulic analysis of core melt relocated in a reactor vessel

International Nuclear Information System (INIS)

Kim, Jong Tae; Kim, Sang Baik; Kim, Hee Dong

2002-03-01

LILAC-meltpool has been developed to study thermo-hydraulic behavior of molten pool and thermal behavior of vessel wall during severe accident. To validate LILAC-meltpool code several two and three dimensional thermo-hydraulic problems were selected and solved. The benchmark problems have experimental results or verified numerical results. Through the validation it was found that LILAC-meltpool reproduces very accurate numerical results. Two-layered semicircular pool was solved to study thermal and hydraulic characteristics of pool stratification. The LAVA experiment using alumina/ferrite molten pool was calculated and compared with computed results. Cooling of alumina/ferrite two-layered pool was affected by stratification. In the numerical results temperature of vessel inner was highest at a location below the interface. Crust was developed from upper surface and lower outer surface, but in the area near the interface corium simulant existed as molten state for long time. LAVA-4 experiment was studied using gap-cooling model in LILAC-meltpool code. Temperature increase of LAVA vessel after alumina melt relocation was strongly dependent on gap formation mechanism. Calculated cooling rates of the vessel were very similar to experimental results. For LAVA experiments which do not have heat generation coolant penetrates easily into a gap and it is found that gap-cooling is very effective for cooling of vessel, but it is thought that coolant penetration could be limited near upper part of gap because of decay heat and high temperature of corium crust

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

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

1994-01-01

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

Test results on direct containment heating by high-pressure melt ejection into the Surtsey vessel: The TDS test series

International Nuclear Information System (INIS)

Allen, M.D.; Blanchat, T.K.; Pilch, M.M.

1994-08-01

The Technology Development and Scoping (TDS) test series was conducted to test and develop instrumentation and procedures for performing steam-driven, high-pressure melt ejection (HPME) experiments at the Surtsey Test Facility to investigate direct containment heating (DCH). Seven experiments, designated TDS-1 through TDS-7, were performed in this test series. These experiments were conducted using similar initial conditions; the primary variable was the initial pressure in the Surtsey vessel. All experiments in this test series were performed with a steam driving gas pressure of ≅ 4 MPa, 80 kg of lumina/iron/chromium thermite melt simulant, an initial hole diameter of 4.8 cm (which ablated to a final hole diameter of ≅ 6 cm), and a 1/10th linear scale model of the Surry reactor cavity. The Surtsey vessel was purged with argon ( 2 ) to limit the recombination of hydrogen and oxygen, and gas grab samples were taken to measure the amount of hydrogen produced

A feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure

International Nuclear Information System (INIS)

Kang, K. H.; Kim, J. H.; Park, L. J.; Kim, S. B.; Hwang, I. S.

1999-01-01

This paper presents the results of a feasibility experiment for assessing the efficacy of ex-vessel cooling through the external gap structure during a severe accident. In this study, a 1/8 linear scale mockup of a lower plenum was used with Al2O3/Fe thermite melt as a corium simulant. The results show that in dry case test conducted without cooling the outside of the vessel, after about thirty second from the thermite ignition the vessel was heated to cause a complete melt penetration at about 30 degree upper position from the bottom. Whereas in wet case test conducted cooling the outside of the vessel with 0.85 kg/s of water flow rate using 2.5 cm of uniform gap structure, the vessel effectively cooled down with 23.7 K/s of cooling rate by nucleate boiling at the surface of the vessel. The results of two-dimensional analyses using FLUENT code show a similar trend of vessel thermal behavior presented in the tests. Synthesized the results of the tests and analyses work, a natural convection of the melt pool could cause the formation of hot spot at the upper portion of the vessel, but the vessel could effectively cool down by heat removal with ex-vessel cooling

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

recommended that further investigations, both experimental and model development, be conducted to (a) check reproducibility of data (b) employ different flow rates (c) employ different simulant materials and (d) develop a comprehensive model, in order to certify that the coolability that can be achieved with establishing a water flow in the CRGTs will be able to retain the melt in the lower head of a BWR. We believe it will be an extremely important accident management strategy for a Swedish BWR since it will obviate the consideration of the prime licensing issue of ex-vessel steam explosion induced containment failure associated with the present scheme of establishing a water pool in the lower drywell of all the Swedish BWRs

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).

36 CFR 13.1180 - Closed waters, motor vessels and seaplanes.

Science.gov (United States)

2010-07-01

... 36 Parks, Forests, and Public Property 1 2010-07-01 2010-07-01 false Closed waters, motor vessels... and Preserve Vessel Operating Restrictions § 13.1180 Closed waters, motor vessels and seaplanes. (a... Hugh Miller Inlet. (4) Waters within the Beardslee Island group (except the Beardslee Entrance), that...

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.)

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

International Nuclear Information System (INIS)

Frolov, Kyrill N.

2003-01-01

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

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

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

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

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-01-01

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

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.)

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)

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)

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)

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

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

43 CFR 423.38 - Operating vessels on Reclamation waters.

Science.gov (United States)

2010-10-01

... 43 Public Lands: Interior 1 2010-10-01 2010-10-01 false Operating vessels on Reclamation waters... WATERBODIES Rules of Conduct § 423.38 Operating vessels on Reclamation waters. (a) You must comply with... Reclamation waters, and with any restrictions established by an authorized official. (b) You must not operate...

Heat insulation device for reactor pressure vessel in water

International Nuclear Information System (INIS)

Nakamura, Heiichiro; Tanaka, Yoshimi.

1993-01-01

Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

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)

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

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.

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.

A water inner circulation device for a reactor vessel

International Nuclear Information System (INIS)

Eriksson, O.

1976-01-01

A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

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

Summary of Reported Whale-Vessel Collisions in Alaskan Waters

Directory of Open Access Journals (Sweden)

Janet L. Neilson

2012-01-01

Full Text Available Here we summarize 108 reported whale-vessel collisions in Alaska from 1978–2011, of which 25 are known to have resulted in the whale's death. We found 89 definite and 19 possible/probable strikes based on standard criteria we created for this study. Most strikes involved humpback whales (86% with six other species documented. Small vessel strikes were most common (<15 m, 60%, but medium (15–79 m, 27% and large (≥80 m, 13% vessels also struck whales. Among the 25 mortalities, vessel length was known in seven cases (190–294 m and vessel speed was known in three cases (12–19 kn. In 36 cases, human injury or property damage resulted from the collision, and at least 15 people were thrown into the water. In 15 cases humpback whales struck anchored or drifting vessels, suggesting the whales did not detect the vessels. Documenting collisions in Alaska will remain challenging due to remoteness and resource limitations. For a better understanding of the factors contributing to lethal collisions, we recommend (1 systematic documentation of collisions, including vessel size and speed; (2 greater efforts to necropsy stranded whales; (3 using experienced teams focused on determining cause of death; (4 using standard criteria for validating collision reports, such as those presented in this paper.

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

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

Melt jet fragmentation and oxidation in the lower plenum

International Nuclear Information System (INIS)

Berthoud, G.

2001-01-01

During the late phases of a PWR Severe Accident, the core materials discharge into the lower plenum in which water is still present. In that case, we are then concerned by the possible occurrence of a Steam Explosion which may endanger the vessel structure and by the following cooling of the melt debris. So, we have two possible ways of vessel rupture: a mechanical one following an energetic Steam Explosion and a thermal one due to insufficient debris cooling. Both types of problems are linked with the degree of fragmentation of the core material during its penetration into the water of the lower plenum. One of the most likely mode of discharge consists in corium streams or jets. The fragmentation will build a corium-water mixture (the pre-mixing sequence) which, under certain circumstances, may undergo a fine fragmentation sequence leading to an energetic Steam Explosion (the explosion sequence). Whatever the occurrence of a Steam Explosion, the resulting debris will accumulate at the bottom of the Reactor Vessel and the cooling of such a ''debris bed'' is known to be highly dependant of the granulometry and build up of the debris bed which are linked with the previous sequence of corium fragmentation and dispersion. In CEA, the MC3D Code has been developed to deal with all these phenomena. (author)

Leakage detecting method and device for water tight vessel of wet-type container apparatus

International Nuclear Information System (INIS)

Tanaka, Yoshimi.

1995-01-01

The present invention provides a method of and a device for detecting leakage of a water tight vessel of a wet-type container apparatus for containing a reactor pressure vessel while immersing it water in a reactor container. Namely, in the wet-type container apparatus, the periphery of the pressure vessel is coated with a heat insulation material and the periphery of the heat insulation material is coated with a water tight vessel. The water tight vessel is immersed under water in the reactor container. As a method of detecting leakage of the wet-type container apparatus, gases mixed with helium are supplied into the water tight vessel at a pressure higher than the inner pressure of the reactor container at a lowest position of the reactor pressure vessel. A water level in the reactor container is determined so as to form a space at the top portion of the inside of the reactor container. The helium at the top portion is detected to monitor the leakage of the water tight vessel. With such procedures, even if the water tight vessel is ruptured at any position, helium mixed to the gases is released to water in the reactor container and rise up to the top space and detected by a helium leakage detection device. (I.S.)

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)

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

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....

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...

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.

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.

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.

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

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)

Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

International Nuclear Information System (INIS)

Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

1983-01-01

The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

Prestressed concrete pressure vessels for boiling water reactors

International Nuclear Information System (INIS)

Menon, S.

1979-12-01

Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

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.

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)

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.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Under Water Thermal Cutting of the Moderator Vessel and Thermal Shield

International Nuclear Information System (INIS)

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

2007-01-01

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

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.

An Experimental Investigation on APR1400 Penetration Weld Failure by Metallic Melt

International Nuclear Information System (INIS)

An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol

2014-01-01

The penetrations are considered as the most vulnerable parts with respect to the reactor vessel failure when a core melt severe accident occurs and the corium reaches the lower head. Penetration tube failure modes can be divided into two categories; tube ejection out of the vessel lower head and rupture of the penetration tube outside the vessel. Tube ejection begins with degrading the penetration tube weld strength to zero as the weld is exposed to temperatures as high as the weld melting temperature, which is called weld failure, and then overcoming any binding force in the hole in the vessel wall that results from differential thermal expansion of the tube and vessel wall. Tube rupture assumes that the debris bed has melted the instrument tube inside the reactor and melt migrates down into the tube to a location outside the vessel wall where a pressure rupture can occur, thus breaching the pressure boundary. In the present paper, we have a focus on the tube ejection failure mode, specifically on the APR1400 weld failure by direct contact with a metallic melt. The objective is to investigate experimentally the ablation kinetics of an APR1400 penetration weld during the interactions with a metallic melt and to suggest the modification of the existing weld failure model. This paper involves the interaction experiments of two different metallic melts (metallic corium and stainless steel melts) with a weld specimen, and rough estimation of weld failure time. The interaction experiments between the metallic melts and an APR1400 penetration weld were performed to investigate the ablation kinetics of the penetration weld. Metallic corium and stainless steel melts were generated using an induction heating technique and interacted with a penetration weld specimen. The ablation rate of the weld specimen showed a range from 0.109 to 0..244 mm/s and thus the APR1400 penetration weld was estimated to be failed at hundreds of times after the interaction with the melt

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

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.

Magnetic resonance imaging of water ascent in embolized xylem vessels of grapevine stem segments

Science.gov (United States)

Mingtao Wang; Melvin T. Tyree; Roderick E. Wasylishen

2013-01-01

Temporal and spatial information about water refilling of embolized xylem vessels and the rate of water ascent in these vessels is critical for understanding embolism repair in intact living vascular plants. High-resolution 1H magnetic resonance imaging (MRI) experiments have been performed on embolized grapevine stem segments while they were...

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.

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)

Experimental and theoretical studies on the high pressure vessel

International Nuclear Information System (INIS)

So, Dong Sup

1992-02-01

A High Pressure Melt Ejection (HPME) is one of the most important phenomena relevant to Direct Containment Heating(DCH) which could lead to an early containment failure in a several accident of PWRs. Dispersal of core debris following a postulated high pressure failure of PWR reactor vessel has been investigated by experimental works and one-dimensional computer modeling to find the relation between the fraction of melt simulant retained in the cavity and the reactor vessel initial conditions as well as to examine the hydrodynamic processes in a reactor cavity geometry. Simulated HPME experiments have been performed with two small-scale (1/25-th and 1/41-st) transparent reactor cavity models of the Young-Gwang unit 1 and 2. Wood's metal and water have been used as melt sumulants while high pressure nitrogen and carbon dioxide have been used as driver gases to simulate the blowdown steam and gas from the breach of the reactor pressure vessel. The high speed movies of the transient tests showed that no fraction of the melt simulant exits the cavity model via the vertical cavity tunnel under its own momentum, and that the discharged simulant from the pressure vessel exits the reactor cavity model during the gas blowdown. The principal removal mechanism seemed to be a combined mechanism of film entrainment and particle levitation due to the driving force of the blowdown gas. Experimental data for the fraction of melt simulant retained in the cavity model (Y f ) during a postulated scenario of the HPME from PWR pressure vessels have been obtained as a function of various test parameters. These data have been used to develop a correlation for Y f that fits all the data (a total of 313 data points) within the standard deviation of 0.054 by means of dimensional analysis and nonlinear least squares optimization technique. The basic effects of important parameters used to describe the HPME accident sequence on the Y f are determined based on the correlation obtained here and

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.

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

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

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

International Nuclear Information System (INIS)

Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W.; Nichols, R.T.; Sweet, D.W.

1991-08-01

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs

Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

Energy Technology Data Exchange (ETDEWEB)

Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

1991-08-01

Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

Test facility for fast gas injections into a vessel filled with water

International Nuclear Information System (INIS)

Wilhelm, D.; Kirstahler, M.

1987-11-01

The Fast Gas Injection Facility (SGI) was set up to study the hydrodynamics during the expansion of a gas bubble into a vessel filled with water. The gas stored in a pressure vessel expands against gravity through a circular duct into a large cylindrical vessel partly with water. This report covers the description of the test facility and the data acquisition. Results of the first test series are added. (orig.) [de

In- and ex-vessel flooding as part of the severe accident strategy in the KERENA reactor

International Nuclear Information System (INIS)

Levi, P.; Fischer, M.

2011-01-01

Currently, AREVA NP is finalizing the basic design of the KERENA reactor, an advanced boiling water reactor with a net electric output of about 1250 MWe. The safety concept in the KERENA reactor is founded on reliable active and passive systems for water supply and heat removal. The passive systems are based on simple physics and do not require operator action. Therefore, a severe accident (SA) with core damage, caused by the subsequent and multiple failures of the safety systems, has an extremely low probability. Despite this, the KERENA design is intended to involve measures that can limit and stop the progression of the severe accident which further reduces the frequency and extent of radioactive releases into the environment. These additional measures include in-vessel and ex-vessel flooding. Flooding is intended to remove the heat from the core or from the reactor pressure vessel (RPV) and transfer it into the containment. There the heat is removed by the active RHR (residual heat removal) system or by the passive CCCs (containment cooling condensers). Both flooding measures are passive and actuated independent of each other by different signals. The study shows that the in-vessel flooding is capable of arresting the core melt progression before a large molten pool can develop. In the unlikely event that the passive in-vessel flooding cannot be actuated or fails, the core will melt and relocate into the lower head of the RPV. In this case, as a further line of defense, decay heat removal can be achieved through the RPV wall into the water in the cavity. In order to assess whether the ex-vessel cooling can ensure RPV wall integrity a dedicated thermodynamics code has been developed which considers heat transfer from the molten corium pool into the RPV wall and the resulting wall ablation. As an input for the code the stratification behavior of the oxidic and metallic phase of the molten pool is examined. In the case of a light metallic phase on top, high heat

Americium behaviour in plastic vessels

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-15

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

Americium behaviour in plastic vessels

International Nuclear Information System (INIS)

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

2010-01-01

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

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.

Status Report on Ex-Vessel Coolability and Water Management

Energy Technology Data Exchange (ETDEWEB)

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

2016-09-15

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

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Americium behaviour in plastic vessels.

Science.gov (United States)

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

2010-01-01

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

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)

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.

Electric melting furnace for waste solidification

International Nuclear Information System (INIS)

Masaki, Toshio.

1990-01-01

To avoid electric troubles or reduction of waste processing performance even when platinum group elements are contained in wastes to be applied with glass solidification. For this purpose, a side electrode is disposed to the side wall of a melting vessel and a central electrode serving as a counter electrode is disposed about at the center inside the melting vessel. With such a constitution, if conductive materials are deposited at the bottom of the furnace or the bottom of the melting vessel, heating currents flow selectively between the side electrode and the central electrode. Accordingly, no electric currents flow through the conductive deposits thereby enabling to prevent abnormal heating in the bottom of the furnace. Further, heat generated by electric supply between the side electrode and the central electrode is supplied efficiently to raw material on the surface of the molten glass liquid to improve the processing performance. Further, disposition of the bottom electrode at the bottom of the furnace enables current supply between the central electrode and the bottom electrode to facilitate the temperature control for the molten glass in the furnace than in the conventional structure. (I.S.)

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

Corrosion of vessel steel during its interaction with molten corium

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Leak isolation self-repairing tape for a water storage vessel and piping against holes

International Nuclear Information System (INIS)

Nagaya, Kosuke; Sekiguchi, Takahiro; Chen, Zhichao; Murakami, Iwanori

2007-01-01

A new type taping for a water storage vessel or piping is presented, in which water leakage is isolated automatically by its self-repairing mechanism against holes. The self-repairing unit (sealant layer) is consisting of three pieces of net with polymer particles inside lattices. Polymer particles, which expand their volume with water, is used for having self-repairing forces. In ordinary tapes, water leaks along the boundary between the tape and the vessel. In order to retain the leak isolation force, this article first discusses a method for making the sealant tape, then develops a method for fixing the sealant to the vessel. The portion of water leakage can be checked on this tape, and the method of detecting the hole or crack portion of the vessel is also presented by using the tape. (author)

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.

46 CFR 178.340 - Stability standards for pontoon vessels on protected waters.

Science.gov (United States)

2010-10-01

... 46 Shipping 7 2010-10-01 2010-10-01 false Stability standards for pontoon vessels on protected waters. 178.340 Section 178.340 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) SMALL....340 Stability standards for pontoon vessels on protected waters. (a) The portion of the deck...

Experimental investigations on vessel-hole ablation during severe accidents

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.; Green, J.A.; Paladino, D.

1997-12-01

This report presents experimental results, and subsequent analyses, of scaled reactor pressure vessel (RPV) failure site ablation tests conducted at the Royal Institute of Technology, Division of Nuclear Power Safety (RIT/NPS). The goal of the test program is to reduce the uncertainty level associated with the phase-change-ablation process, and, thus, improve the characterization of the melt discharge loading on the containment. In a series of moderate temperature experiments, the corium melt is simulated by the binary oxide CaO-B 2 O 3 or the binary eutectic and non-eutectic salts NaNO 3 -KNO 3 , while the RPV head steel is represented by a Pb, Sn or metal alloys plate. A complementary set of experiments was conducted at lower temperatures, using water as melt and salted ice as plate material. These experiments scale well to the postulated prototypical conditions. The multidimensional code HAMISA, developed at RIT/NPS, is employed to analyze the experiments with good pre- and post-test predictions. The effects of melt viscosity and crust surface roughness, along with failure site entrance and exit frictional losses on the ablation characteristics are investigated. Theoretical concept was proposed to describe physical mechanisms which govern the vessel-hole ablation process during core melt discharge from RPV. Experimental data obtained from hole ablation tests and separate-effect tests performed at RIT/NPS were used to validate component physical models of the HAMISA code. It is believed that the hole ablation phenomenology is quite well understood. Detailed description of experiments and experimental data, as well as results of analyses are provided in the appendixes

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

Assessment of reactor vessel integrity (ARVI)

Energy Technology Data Exchange (ETDEWEB)

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

2003-04-01

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

Effect of sunlight, transport and storage vessels on drinking water ...

African Journals Online (AJOL)

Effect of sunlight, transport and storage vessels on drinking water quality in rural Ghana. ... on drinking water quality in rural Ghana. K Obiri-Danso, E Amevor, LA Andoh, K Jones ... Full Text: EMAIL FREE FULL TEXT EMAIL FREE FULL TEXT

Reactor vessel pressure transient protection for pressurized water reactors

International Nuclear Information System (INIS)

Zech, G.

1978-09-01

During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

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

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)

Offshore support vessel developments for deep water oil and gas E and P

Energy Technology Data Exchange (ETDEWEB)

Dielen, Baldo A.M. [SMIT, Rotterdam (Netherlands)

2008-07-01

The worldwide trend to move towards more exposed locations and deeper waters for O and G exploration and production activities resulted in an increased need for larger and more powerful tugs and offshore support vessels. These vessels must meet higher operational requirements under higher wind and sea-state conditions. This market-driven need, together with technological developments, is leading towards a new generation of powerful and sophisticated offshore support vessels (OSV's). This paper will describe the actual and future trends in OSV design for deep water offshore use. (author)

Structural failure analysis of reactor vessels due to molten core debris

International Nuclear Information System (INIS)

Pfeiffer, P.A.

1993-01-01

Maintaining structural integrity of the reactor vessel during a postulated core melt accident is an important safety consideration in the design of the vessel. This paper addresses the failure predictions of the vessel due to thermal and pressure loadings from the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on a wet or dry cavity and pressurization of the vessel based on operating pressure or atmospheric (pipe break). The analyses considered both short term (minutes) and long term (days) failure modes. Short term failure modes include creep at elevated temperatures and plastic instabilities of the structure. Long term failure modes are caused by creep rupture that lead to plastic instability of the structure. The analyses predict the reactor vessel will remain intact after the core melt has deposited on the lower vessel head

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

Ex-vessel water-level and fission-product monitoring for LWR

International Nuclear Information System (INIS)

DeVolpi, A.; Markoff, D.

1988-01-01

Given that the need for direct measurement of reactor coolant inventory under operational or abnormal conditions remains unsatisfied, a high-energy gamma-ray detection system is described for ex-vessel monitoring. The system has been modeled to predict response in a PWR, and the model has been validated with a LOFT LOCA sequence. The apparatus, situated outside the pressure vessel, would give relative water level and density over the entire vessel height and distinguish differing levels in the downcomer and core. It would also have significant sensitivity after power shutdown because of high-energy gamma rays from photoneutron capture, the photoneutrons being the result of fission-product decay in the core. Fission-products released to the coolant and accumulated in the top of a PWR vessel would also be theoretically detectable

Research on water hammer forces caused by rapid growth of bubbles at severe accidents of water cooled reactors

International Nuclear Information System (INIS)

Inasaka, Fujio; Adachi, Masaki; Aya, Izuo

2004-01-01

At severe accidents of Water Cooled Reactors a great deal of gas is expected to be produced in a short time within the water of lower part of nuclear pressure vessel and containment vessel caused by hydrogen production with a metal water reaction and steam explosions with direct contact of melting core and water. Water hammer forces caused by rapid growth of bubbles shall work on the wall of containment vessel and affect its integrity. Coherency of water block movement is not clear, whether simultaneous or in the same direction. Water block behavior and water hammer forces caused by rapid growth of bubbles have been tested using a modified scale model and analyzed to obtain experimental correlated equation to estimate water block's rising distance and velocity from water hammer data. Numerical analysis using RELAP5-3D (Reactor Excursion and Leak Analysis Program) has been conducted to evaluate water hammer forces and makes clear its modifications needed. (T. Tanaka)

Investigation of the failure of a reactor pressure vessel by plastic instability

International Nuclear Information System (INIS)

Laemmer, H.; Ritter, B.

1994-01-01

A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

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.

The water-filled versus air-filled status of vessels cut open in air: the 'Scholander assumption' revisited

Science.gov (United States)

M.T. Tyree; H. Cochard; P. Cruziat

2003-01-01

When petioles of transpiring leaves are cut in the air, according to the 'Scholander assumption', the vessels cut open should fill with air as the water is drained away by continued transpiration, The distribution of air-filled vessels versus distance from the cut surface should match the distribution of lengths of 'open vessels', i.e. vessels cut...

Principles of application of mechanical design measures to control severe accident phenomena, applied to the melt retention concept of the EPR

International Nuclear Information System (INIS)

Bittermann, D.

2000-01-01

To retain and stabilize a core melt within the containment, the phenomena which principally have to be dealt with are related to melt discharge, spreading, retention and cooling, plus specific phenomena like melt dispersal and ex-vessel melt water interaction. For the elaboration of mechanical design measures provided to stabilize a melt within the containment, boundary conditions may occur which could pose extremely high thermal and mechanical loads on the structures. This file describes an approach characterized by the idea to influence the course of severe accident scenarios as much as possible in order to generate boundary conditions for mitigation means ''by design'', which enables the development of a mitigation concept with maximum confidence in the effectiveness of the measures provided. (orig.)

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)

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.

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.

Detection and characterization of flaws in segments of light water reactor pressure vessels

International Nuclear Information System (INIS)

Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

1988-01-01

Studies have been conducted to determine flaw density in segments cut from light water reactor )LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (H SST) Program. Segments from the Hope Creek Unit 2 vessel and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication [with a through-wall dimension of ∼6 mm (∼0.24 in.)] was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications [i.e., for a total of approximately 6.8 m 2 (72 ft 2 ) of cladding surface]. (author)

77 FR 58213 - Requested Administrative Waiver of the Coastwise Trade Laws: Vessel ISLAND WATERS; Invitation for...

Science.gov (United States)

2012-09-19

... DEPARTMENT OF TRANSPORTATION Maritime Administration [Docket No. MARAD 2012 0095] Requested Administrative Waiver of the Coastwise Trade Laws: Vessel ISLAND WATERS; Invitation for Public Comments AGENCY... ISLAND WATERS is: Intended Commercial Use of Vessel: ``Charter to individuals in coastal waters...

Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

International Nuclear Information System (INIS)

Borovkov, A.I.; Semenov, A.S.; Granovsky, V.S.; Kovtunova, S.V.

1995-01-01

The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs

Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

Energy Technology Data Exchange (ETDEWEB)

Borovkov, A.I.; Semenov, A.S. [St. Petersburg State Technical Univ. (Russian Federation); Granovsky, V.S.; Kovtunova, S.V. [Research Inst. of Technology, Sosnovy Bor (Russian Federation)

1995-12-31

The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs.

Debris interactions in reactor vessel lower plena during a severe accident. II. Integral analysis

International Nuclear Information System (INIS)

Suh, K.Y.; Henry, R.E.

1996-01-01

For pt.I see ibid., p.147-63, 1996. The integral physico-numerical model for the reactor vessel lower head response has been exercised for the TMI-2 accident and possible severe accident scenarios in PWR and BWR designs. The proposed inherent cooling mechanism of the reactor material creep and subsequent water ingression implemented in this predictive model provides a consistent representation of how the debris was finally cooled in the TMI-2 accident and how the reactor lower head integrity was maintained during the course of the incident. It should be recalled that in order for this strain to occur, the vessel lower head had to achieve temperatures in excess of 1000 C. This is certainly in agreement with the temperatures determined by metallographic examinations during the TMI-2 vessel inspection program. The integral model was also applied to typical PWR and BWR lower plena with and without structures under pressurized conditions spanning the first relocation of core material to the reactor vessel failure due to creep without recovery actions. The design application results are presented with particular attention being focused on water ingression into the debris bed through the gap formed between the debris and the vessel wall. As an illustration of the accident management application, the lower plenum with structures was recovered after an extensive amount of creep had damaged the vessel wall. The computed lower head temperatures were found to be significantly lower (by more than 300 K in this particular example) with recovery relative to the case without recovery. This clearly demonstrates the potential for in-vessel cooling of the reactor vessel without a need to externally submerge the lower head should such a severe accident occur as core melting and relocation. (orig.)

33 CFR 165.1411 - Security zone; waters surrounding U.S. Forces vessel SBX-1, HI.

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Security zone; waters surrounding U.S. Forces vessel SBX-1, HI. 165.1411 Section 165.1411 Navigation and Navigable Waters COAST GUARD... § 165.1411 Security zone; waters surrounding U.S. Forces vessel SBX-1, HI. (a) Location. The following...

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)

Problem of corium melt coolability in passive protection systems against severe accidents in the containment

Directory of Open Access Journals (Sweden)

Ali Kalvand

2018-05-01

Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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.

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Phenomena in the interaction among a core melt and protective and sacrificial materials

International Nuclear Information System (INIS)

Steinwarz, W.; Koller, W.; Dyllong, N.; Fischer, M.; Hellmann, S.; Lansmann, V.; Nie, M.; Haefner, W.; Alkan, Z.; Andrae, P.; Rensing, B.

2000-01-01

In a postulated core meltdown accident in a light water reactor there are bound to be interactions, in the ex-vessel phase, among the core melt and the structural materials within and below the reactor cavity. In existing plants, these structural materials normally are structural concrete, while future, evolutionary reactor lines are to have sacrificial and protective materials specially designed for this hypothetical case. To add to the state of knowledge about the phenomena occurring, experiments need to be conducted under conditions as realistic as possible. Within the research programs funded by the European Union, the German Federal Ministry for Economics, and the German nuclear power plant operators, experiments on a laboratory as well as an industrial scale on these problems are being carried out in the two projects called CORESA (COrium on REfractory and SAcrificial materials) and ECOSTAR (Ex-vessel COre melt STAbilization Research). The experiments are accompanied by an extensive analytical theoretical program also serving to advance and validate computer codes on the problems under investigation. The projects, which are carried out with international European participation, are expected to allow a concept to be developed for managing postulated accident scenarios involving core meltdown for innovative nuclear power plants, and to provide findings on risk evaluation of plants now in operation so as to further develop accident management measures. (orig.) [de

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

The coolability limits of a reactor pressure vessel lower head

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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...

Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

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

1991-01-01

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

Experimental Study of Air Vessel Behavior for Energy Storage or System Protection in Water Hammer Events

Directory of Open Access Journals (Sweden)

Mohsen Besharat

2017-01-01

Full Text Available An experimental assessment of an air pocket (AP, confined in a compressed air vessel (CAV, has been investigated under several different water hammer (WH events to better define the use of protection devices or compressed air energy storage (CAES systems. This research focuses on the size of an AP within an air vessel and tries to describe how it affects important parameters of the system, i.e., the pressure in the pipe, stored pressure, flow velocity, displaced volume of water and water level in the CAV. Results present a specific range of air pockets based on a dimensionless parameter extractable for other real systems.

Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

International Nuclear Information System (INIS)

Thakre, S.; Ma, W.

2013-08-01

Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

Simulations of ex-vessel fuel coolant interactions in a Nordic BWR using MC3D code

Energy Technology Data Exchange (ETDEWEB)

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

2013-08-15

Nordic Boiling Water Reactors (BWRs) employ a drywell cavity flooding technique as a nuclear severe accident management strategy. In case of core melt accident where the reactor pressure vessel will fail and the melt will eject from the lower head and fall into a water pool, may be in the form of a continuous jet. It is assumed that the melt jet will fragment, quench and form a coolable debris bed into the water pool. The melt interaction with a water pool may cause an energetic steam explosion which creates a potential risk towards the integrity of containment, leading to fission products release into the atmosphere. The results of the APRI-7 project suggest that the significant damage to containment structures by steam explosion cannot be ruled according to the state-of-the-art knowledge about corresponding accident scenario. In the follow-up project APRI-8 (2012-2016) one of the goals of the KTH research is to resolve the steam explosion energetics (SEE) issue, developing a risk-oriented framework for quantifying conditional threats to containment integrity for a Nordic type BWR. The present study deals with the premixing and explosion phase calculations of a Nordic BWR dry cavity, using MC3D, a multiphase CFD code for fuel coolant interactions. The main goal of the study is the assessment of pressure buildup in the cavity and the impact loading on the side walls. The conditions for the calculations are used from the SERENA-II BWR case exercise. The other objective was to do the sensitivity analysis of the parameters in modeling of fuel coolant interactions, which can help to reduce uncertainty in assessment of steam explosion energetics. The results show that the amount of liquid melt droplets in the water (region of void<0.6) is maximum even before reaching the jet at the bottom. In the explosion phase, maximum pressure is attained at the bottom and the maximum impulse on the wall is at the bottom of the wall. The analysis is carried out using two different

Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2010-01-01

1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

An Approach for Selection of Flow Regime and Models for Conservative Evaluation of a Vessel Integrity Monitoring System for Water-Cooled Vacuum Vessels

International Nuclear Information System (INIS)

Pointer, W. David; Ruggles, Arthur E.

2003-01-01

Thin-walled vacuum containment vessels cooled by circulating water jackets are often utilized in research and industrial applications where isolation of equipment or experiments from the influences of the surrounding environment is desirable. The development of leaks in these vessels can result in costly downtime for the facility. A Vessel Integrity Monitoring System (VIMS) is developed to detect leak formation and estimate the size of the leak to allow evaluation of the risk associated with continued operation. A wide range of leak configurations and fluid flow phenomena are considered in the evaluation of the rate at which a tracer gas dissolved in the cooling jacket water is transported into the vacuum vessel. A methodology is presented that uses basic fluid flow models and careful evaluation of their ranges of applicability to provide a conservative estimate of the transport rates for the tracer gas and hence the time required for the VIMS to detect a leak of a given size

Coupled thermo-mechanical analysis of corium-loaded lower head of pressure vessel

International Nuclear Information System (INIS)

Mishra, J.; Balasubramaniyan, V.

2016-01-01

A severe accident in the pressurised water reactor may lead to the relocation of core materials to the lower head of Reactor Pressure Vessel (RPV). The core debris at the bottom of RPV forms a melt pool of corium due to decay heat. The understanding of behaviour of pressure vessel, characterised by failure mode and time to failure, in this scenario is one of the important steps in predicting the accident progression. The most predominant failure mode is multi-axial creep deformation of the vessel with a non-uniform temperature field. Towards this, a numerical analysis methodology is developed for the prediction of pressure vessel deformation during the severe accidents. The methodology involves 2-D finite element modelling under multi-physics environment, which account the creep phenomena using Norton-Bailey creep law with a typical damage model of RPV material. The validation of the methodology is carried out using the results from OLHF experiment carried out in Sandia National Laboratory (SNL), USA, within the framework of an OECD. (author)

Multiple shell pressure vessel

International Nuclear Information System (INIS)

Wedellsborg, B.W.

1988-01-01

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

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.

Head spray nozzle in reactor pressure vessel

International Nuclear Information System (INIS)

Hatano, Shun-ichi.

1990-01-01

In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

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

Conjunctival-corneal melt in association with carotid artery stenosis

Directory of Open Access Journals (Sweden)

Rosalind MK Stewart

2008-10-01

Full Text Available Rosalind MK Stewart1, Say Aun Quah1, Dan Q Nguyen2, Stephen B Kaye11Royal Liverpool University Hospital, Liverpool, UK; 2Bristol Eye Hospital, Bristol, UKPurpose: To report a case of severe conjunctival-corneal melt in association with carotid artery stenosis.Methods: Observational case report.Results: A 76-year-old man with a history of bilateral severe carotid artery occlusion and nonarteritic ischemic optic neuropathy developed a spontaneous bulbar conjunctival defect. Despite intensive lubrication, and attempts at surgical closure including an amniotic membrane patch graft, it progressed with subsequent adjacent corneal perforation. Thorough investigations revealed no underlying disease, except markedly delayed episcleral vessel filling on anterior segment fluorescein angiography.Conclusions: Neovascularisation is a known factor in the inhibition of ulceration. In light of the findings in this report, ocular ischemia should be considered as a cause or contributing factor in the differential diagnosis of conjunctival-corneal melt.Keywords: conjunctival melt, corneal melt, ocular ischemia, carotid artery stenosis

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.

Interaction between molten corium UO{sub 2+x}-ZrO{sub 2}-FeO{sub y} and VVER vessel steel

Energy Technology Data Exchange (ETDEWEB)

Bechta, S. V.; Granovsky, V. S.; Khabensky, V. B.; Krushinov, E. V.; Vitol, S. A.; Sulatsky, A. A. [Alexandrov Sci Res Technol Inst, Sosnovyi Bor (Russian Federation); Gusarov, V. V.; Almiashev, V. I. [Russian Acad Sci, Inst Silicate Chem, St Petersburg (Russian Federation); Lopukh, D. B. [SPb State Electrotech Univ LETI SPbGETU, St Petersburg (Russian Federation); Bottomley, D. [Joint Res Ctr, Inst Transurane, Karlsruhe (Germany); Fischer, M. [AREVA NP GmbH, Erlangen (Germany); Piluso, P. [CEA Saclay, DEN, DSNI, Saclay (France); Miassoedov, A.; Tromm, W. [Forschungszentrum Karlsruhe, D-76021 Karlsruhe (Germany); Altstadt, E. [Forschungszentrum Dresden Rossendorf, Dresden (Germany); Fichot, F. [CEA Cadarache, SEMCA, DPAM, IRSN, St Paul Les Durance (France); Kymalainen, O. [FORTUM Nucl Serv Ltd, Espoo (Finland)

2010-07-01

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

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

Standard Guide for In-Service Annealing of Light-Water Moderated Nuclear Reactor Vessels

CERN Document Server

American Society for Testing and Materials. Philadelphia

2003-01-01

1.1 This guide covers the general procedures to be considered for conducting an in-service thermal anneal of a light-water moderated nuclear reactor vessel and demonstrating the effectiveness of the procedure. The purpose of this in-service annealing (heat treatment) is to improve the mechanical properties, especially fracture toughness, of the reactor vessel materials previously degraded by neutron embrittlement. The improvement in mechanical properties generally is assessed using Charpy V-notch impact test results, or alternatively, fracture toughness test results or inferred toughness property changes from tensile, hardness, indentation, or other miniature specimen testing (1). 1.2 This guide is designed to accommodate the variable response of reactor-vessel materials in post-irradiation annealing at various temperatures and different time periods. Certain inherent limiting factors must be considered in developing an annealing procedure. These factors include system-design limitations; physical constrain...

Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

International Nuclear Information System (INIS)

Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

1980-01-01

An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

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)

Corrosion of vessel steel during its interaction with molten corium

International Nuclear Information System (INIS)

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

2006-01-01

An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments

Corrosion of vessel steel during its interaction with molten corium

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation)]. E-mail: bechta@sbor.spb.su; Khabensky, V.B. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Vitol, S.A. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Krushinov, E.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Granovsky, V.S. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Lopukh, D.B. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Gusarov, V.V. [Institute of Silicate Chemistry of Russian Academy of Science (ISC of RAS), Odoevsky str., b. 24/2, 199155 St. Petersburg (Russian Federation); Martinov, A.P. [SPb Electrotechnical University (SpbGETU), Professor Popov str., b.5/3, 197376 St. Petersburg (Russian Federation); Martinov, V.V. [Scientific Research Technological Institute (NITI), Sosnovy Bor of Leningrad Oblast 188540 (Russian Federation); Fieg, G. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Tromm, W. [Forshungszentrum Karlsruhe (FZK), Institut fur Neutronenphysik and Reaktortechnik, Postfach 3640, D-78021 Karlsruhe (Germany); Bottomley, D. [Europaeische Kommission, General Direktion GFS, Institut fuer Transurane (ITU), Postfach 2340, 76125 Karlsruhe (Germany); Tuomisto, H. [Fortum Engineering Ltd. 00048 FORTUM, Rajatorpantie 8, Vantaa (Finland)

2006-07-15

An experimental examination of the cooled vessel steel corrosion during the interaction with molten corium is presented. The experiments have been conducted on 'Rasplav-2' test facility and followed up with physico-chemical and metallographic analyses of melt samples and corium-specimen ingots. The results discussed in the first part of the paper have revealed specific corrosion mechanisms for air and inert atmosphere above the melt. Models have been proposed based on this information and approximate curves constructed for the estimation of the corrosion rate or corrosion depth of vessel steel in conditions simulated by the experiments.

A study on corium melt pool behavior under external vessel cooling : investigation of the first phase research results in the OECD RASPLAV project

Energy Technology Data Exchange (ETDEWEB)

Park, Rae Joon; Kim, Sang Baik; Kim, Hee Dong; Yoo, Kun Joong

1998-04-01

The scope and contents of the OECD RASPLAV program are to investigate natural convection heat transfer in the corium, chemical and mechanical interaction between the corium and the reactor vessel, crust formation of the corium, and thermal behaviour of the corium by experiments and model development during external vessel cooling to prevent reactor vessel failure in severe accidents of nuclear power plant. This study includes evaluation and analysis of the RASPLAV V phase I results for three years between July 1, 1994 and June 30, 1997. These results supply technical basis for our experimental program on severe accident research. Two large-scale experiments of RASPLAV-AW-between the corium and the reactor vessel. Several small-scale experiments were conducted to analyze thermal stratification in the corium. The salt experiments were conducted to estimate the crust and the mushy region formation, and natural convection heat transfer in the corium. In the analytical studies, pre and post analysis of the RASPLAV-AW-200 experiments and evaluation of the salt test results have been performed using CONV 2 and 3D computer codes, which were developed during RASPLAV program phase I. Low density corium was separated from the high density corium during the RASPLAV-AW-200 tests and the TULPAN test, which was a new finding in the RASPLAV project phase I. From the salts test, heat flux distribution in the side wall heating case is similar to the direct internal heat generation case, and the crust formation is a little effect on heat transfer rate. The results of CONV 2 and 3 D were very well with with the experimental results. The results of RASLAV project phase I, such as furnace design and the techniques on fuel melting, are very helpful to our severe accident experimental program. (author). 57 refs., 13 tabs., 52 figs.

Development of high-frequency induction melting system for radioactive solid wastes

International Nuclear Information System (INIS)

Kawaguchi, Ichiro; Yamazaki, Seichiro; Takahashi, Noriaki; Kugai, Katsutoshi; Yokozawa, Minoru

2004-01-01

Kawasaki Heavy Industries, Ltd. developed an active insulation (AI) method radiofrequency melting system as a new melting treatment system of radioactive solid wastes and proved production of waste satisfied the treatment performances and burying by repeating many practical melting tests. The melting vessel uses a low-priced ceramic canister with nonelectrical conductivity, which is able to treat wastes with large amount of inorganic substances. The wastes melted in the canister is taken out the canister itself from radiofrequency melting reactor and solidified after cooling. The cool canister is stored in 2001 metal drum filling up a gap with mortal for laying underground. New radiofrequency melting reactor, 1/3 scale melting test, estimation of scale effects, melting tests for practical use and the total system are explained. (S.Y.)

Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

International Nuclear Information System (INIS)

Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

2017-01-01

Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

A Self-Assembled Trigonal Prismatic Molecular Vessel for Catalytic Dehydration Reactions in Water.

Science.gov (United States)

Das, Paramita; Kumar, Atul; Howlader, Prodip; Mukherjee, Partha Sarathi

2017-09-12

A water-soluble Pd 6 trigonal prism (A) was synthesized by two-component coordination-driven self-assembly of a Pd II 90° acceptor with a tetraimidazole donor. The walls of the prism are constructed by three conjugated aromatic building blocks, which means that the confined pocket of the prism is hydrophobic. In addition to the hydrophobic cavity, large product egress windows make A an ideal molecular vessel to catalyze otherwise challenging pseudo-multicomponent dehydration reactions in its confined nanospace in aqueous medium. This study is an attempt at selective generation of the intermediate tetraketones and xanthenes by fine-tuning the reaction conditions employing a supramolecular molecular vessel. Moreover, either poor or no yield of the dehydrated products in the absence of A under similar reaction conditions supports the ability of the confined space of the barrel to promote such reactions in water. Furthermore, we focused on the rigidification of the tetraphenylethylene-based tetraimidazole unit anchored within the Pd II coordination architecture; enabling counter-anion dependent aggregation induced emission in the presence of water. © 2017 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

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.)

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...

Method and device for catching reactor core melt-down masses in hypothetical accidents of nuclear power plants

International Nuclear Information System (INIS)

Morlock, G.; Wiesemes, J.; Bachner, D.

1977-01-01

The device is to receive the afterheat of the molten core and in this way to prevent afterflow of coolant and a new criticality. A tank below the reactor pressure vessel, with the proper diameter, contains a store of salt or a salt mixture suitable to receive the afterheat of a core melt-down as heat of fusion or conversion. Above the salt, there is a layer of thermoplastics or of a material forming a hardening foam. Coolant eventually continuing to flow out is separated from the core melt by this barrier layer, and thus the build-up of high steam pressures is prevented. Neutron-absorbing materials, like boron salts mixed to the salts, as well as a subdivision of the salt surface, e.g. by means of canalizing firebricks, prevent the formation of new criticality. Further installations within the tank, like pipings or channels, permit the introduction of water after cooling down of the core or salt melt-down mass and to wash out the brine with all radioactive and other constituents for transport to reprocessing or ultimate storage. (HP) [de

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

International Nuclear Information System (INIS)

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

1995-09-01

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

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)

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.

Experimental and numerical study of water-filled vessel impacted by flat projectiles

Science.gov (United States)

Zhang, Wei; Ren, Peng; Huang, Wei; Gao, Yu Bo

2014-05-01

To understand the failure modes and impact resistance of double-layer plates separated by water, a flat-nosed projectile was accelerated by a two-stage light gas gun against a water-filled vessel which was placed in an air-filled tank. Targets consisted of a tank made of two flat 5A06 aluminum alloy plates held by a high strength steel frame. The penetration process was recorded by a digital high-speed camera. The same projectile-target system was also used to fire the targets placed directly in air for comparison. Parallel numerical tests were also carried out. The result indicated that experimental and numerical results were in good agreement. Numerical simulations were able to capture the main physical behavior. It was also found that the impact resistance of double layer plates separated by water was lager than that of the target plates in air. Tearing was the main failure models of the water-filled vessel targets which was different from that of the target plates in air where the shear plugging was in dominate.

Considerations concerning the strategy of corium retention in the reactor vessel

International Nuclear Information System (INIS)

2015-01-01

Third-generation nuclear reactors are characterised by consideration during design of core meltdown accidents. More specifically, dedicated measures or devices must be implemented to avoid basemat melt-through in the reactor building. These devices must have a high level of confidence. The strategy of corium retention in the reactor vessel, if supported by appropriate research and development, makes it possible to achieve this objective. IRSN works alone or in partnerships to address all the issues associated with in-vessel corium retention. This document describes the in-vessel corium retention strategy and its limitations, along with the research programs conducted by IRSN in this area

Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

International Nuclear Information System (INIS)

Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

1977-01-01

After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)

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.

OECD MCCI project Melt Eruption Test (MET) design report, Rev. 2. April 15, 2003

International Nuclear Information System (INIS)

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

2011-01-01

satisfy these PRG recommendations. Specifically, the revised plan focuses on providing data on the extent of crust growth and melt eruptions as a function of gas sparging rate under well-controlled experiment conditions, including a floating crust boundary condition. The overall objective of MET is to determine to what extent core debris is rendered coolable by eruptive-type processes that breach the crust that rests upon the melt. The specific objectives of this test are as follows: (1) Evaluate the augmentation in surface heat flux during periods of melt eruption; (2) Evaluate the melt entrainment coefficient from the heat flux and gas flow rate data for input into models that calculate ex-vessel debris coolability; (3) Characterize the morphology and coolability of debris resulting from eruptive processes that transport melt into overlying water; and (4) Discriminate between periods when eruptions take the form of particle ejections into overlying water, leading to a porous particle bed, and single-phase extrusions, which lead to volcano-type structures.

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-06-15

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1982-01-01

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

Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

International Nuclear Information System (INIS)

Kuroda, Hidehiko; Okazaki, Koki; Shiraishi, Fujio; Kenjyo, Hiroaki; Isoda, Koichiro

2013-01-01

TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

Monitoring device for glass melting furnace

International Nuclear Information System (INIS)

Endo, Noboru; Asano, Naoki; Higuchi, Tatsuo; Koyama, Mayumi; Hanado, Shinji.

1995-01-01

The device of the present invention can monitor, from a remote place, a liquid surface in a glass melting furnace for use in a solidification treatment, for example, of high level radioactive wastes. Namely, a vertical sleeve is disposed penetrating a ceiling wall of a melting vessel. A reflection mirror is disposed above the vertical sleeve and flex an optical axis. A monitoring means is disposed on the optical axis of the reflecting mirror at a spaced position. The monitoring means may have an optical telescopic means, a monitoring camera by way of a half mirror and an illumination means. The reflection mirror may be made of a metal. The monitoring device thus constituted suffer from no effects of high temperature and high radiation dose rate, thereby enabling to easily monitor the liquid surface in the melting furnace. (I.S.)

A fast running method for predicting the efficiency of core melt spreading for application in ASTEC

International Nuclear Information System (INIS)

Spengler, C.

2010-01-01

The integral Accident Source Term Evaluation Code (ASTEC) is jointly developed by the French Institut de Radioprotection et de Surete Nucleaire (IRSN) and the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH to simulate the complete scenario of a hypothetical severe accident in a nuclear light water reactor, from the initial event until the possible radiological release of fission products out of the containment. In the frame of the new series of ASTEC V2 versions appropriate model extensions to the European Pressurised Water Reactor (EPR) are under development. With view to assessing with ASTEC the proper operation of the ex-vessel melt retention and coolability concept of the EPR with regard to melt spreading an approximation of the area finally covered by the corium and of the distance run by the corium front before freezing is required. A necessary capability of ASTEC is in a first step to identify such boundary cases, for which there is a potential that the melt will freeze before the spreading area is completely filled. This paper presents a fast running method for estimating the final extent of the area covered with melt on which a simplified criterion in ASTEC for detecting such boundary cases will be based. If a boundary case is detected the application of a more-detailed method might be necessary to assess further the consequences for the accident sequence. The major objective here is to provide a reliable method for estimating the final result of the spreading and not to provide highly detailed methods to simulate the dynamics of the transient process. (orig.)

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.

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.

Hydrostatic extrusion of Cu-Ag melt spun ribbon

Science.gov (United States)

Hill, Mary Ann; Bingert, John F.; Bingert, Sherri A.; Thoma, Dan J.

1998-01-01

The present invention provides a method of producing high-strength and high-conductance copper and silver materials comprising the steps of combining a predetermined ratio of the copper with the silver to produce a composite material, and melt spinning the composite material to produce a ribbon of copper and silver. The ribbon of copper and silver is heated in a hydrogen atmosphere, and thereafter die pressed into a slug. The slug then is placed into a high-purity copper vessel and the vessel is sealed with an electron beam. The vessel and slug then are extruded into wire form using a cold hydrostatic extrusion process.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

Development of a surrogate model for analysis of ex-vessel steam explosion in Nordic type BWRs

Energy Technology Data Exchange (ETDEWEB)

Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Basso, Simone, E-mail: simoneb@kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2016-12-15

Highlights: • Severe accident. • Steam explosion. • Surrogate model. • Sensitivity study. • Artificial neural networks. - Abstract: Severe accident mitigation strategy adopted in Nordic type Boiling Water Reactors (BWRs) employs ex-vessel core melt cooling in a deep pool of water below reactor vessel. Energetic fuel–coolant interaction (steam explosion) can occur during molten core release into water. Dynamic loads can threaten containment integrity increasing the risk of fission products release to the environment. Comprehensive uncertainty analysis is necessary in order to assess the risks. Computational costs of the existing fuel–coolant interaction (FCI) codes is often prohibitive for addressing the uncertainties, including the effect of stochastic triggering time. This paper discusses development of a computationally efficient surrogate model (SM) for prediction of statistical characteristics of steam explosion impulses in Nordic BWRs. The TEXAS-V code was used as the Full Model (FM) for the calculation of explosion impulses. The surrogate model was developed using artificial neural networks (ANNs) and the database of FM solutions. Statistical analysis was employed in order to treat chaotic response of steam explosion impulse to variations in the triggering time. Details of the FM and SM implementation and their verification are discussed in the paper.

Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

International Nuclear Information System (INIS)

Li, D D; Jiang, J; Zhao, Z; Yi, W S; Lan, G

2013-01-01

We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system

Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

Science.gov (United States)

Li, D. D.; Jiang, J.; Zhao, Z.; Yi, W. S.; Lan, G.

2013-12-01

We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system.

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Investigating plume dynamics at the ocean-glacier interface with turbulence profiling and autonomous vessels

Science.gov (United States)

Jackson, R. H.; Nash, J. D.; Sutherland, D. A.; Amundson, J. M.; Kienholz, C.; Skyllingstad, E. D.; Motyka, R. J.

2017-12-01

The exchanges of heat and freshwater at tidewater glacier termini are modulated by small-scale turbulent processes. However, few observations have been obtained near the ocean-glacier interface, limiting our ability to quantify turbulent fluxes or test melt parameterizations in ocean-glacier models. Here, we explore the turbulent plume dynamics at LeConte Glacier, Alaska with three extensive field campaigns in May, August and September (2016-17). Two autonomous vessels collected repeat transects of velocity and water properties near the glacier, often within 20 m of the terminus. Concurrent shipboard surveying measured turbulence with a vertical microstructure profiler, along with water properties and velocity. These high-resolution surveys provide a 3D view of the circulation and allow us to quantify turbulent fluxes in the near-glacier region. We observe two regimes at the terminus: an energetic upwelling plume driven by subglacial discharge at a persistent location, and submarine melt-driven convection along other parts of the terminus. We trace the evolution of the subglacial discharge plume as it flows away from the glacier, from an initial stage of vigorous mixing to a more quiescent outflow downstream. Resolving these spatial patterns of upwelling and mixing near glaciers is a key step towards understanding submarine melt rates and glacial fjord circulation.

Analytical and Experimental Study for Validation of the Device to Confine BN Reactor Melted Fuel

International Nuclear Information System (INIS)

Rogozhkin, S.; Osipov, S.; Sobolev, V.; Shepelev, S.; Kozhaev, A.; Mavrin, M.; Ryabov, A.

2013-01-01

To validate the design and confirm the design characteristics of the special retaining device (core catcher) used for protection of BN reactor vessel in the case of a severe beyond-design basis accident with core melting, computational and experimental studies were carried out. The Tray test facility that uses water as coolant was developed and fabricated by OKBM; experimental studies were performed. To verify the methodical approach used for the computational study, experimental results obtained in the Tray test facility were compared with numerical simulation results obtained by the STAR-CCM+ CFD code

Vessel for solidifying water-impermeable radioactive waste

International Nuclear Information System (INIS)

Kiuchi, Yoshimasa; Tamada, Shin; Suzuki, Yasushi.

1993-01-01

A blend prepared by admixing silica sand, alumina powder or glass fiber, as aggregates, to epoxy resin elastic adhesives is coated on an inner surface of a steel drum can or an inner surface of a concrete vessel at a thickness of greater than 1mm followed by hardening. The addition amount of the silica sand, alumina powder or glass fiber is determined as 20 to 40% by weight, 30 to 60% by weight or 5 to 15% by weight respectively. A lid having a hole for injecting fillers is previously bonded to a container for use in solidifying radioactive materials. The strength of the coating layer is increased and a coating performance and an adhesion force are improved by admixing the aggregates, to provide a satisfactory water-impermeability. The container for use in solidifying radioactive wastes having a coating layer with an advantage of the elastic resin adhesives, strong strength and adhesion and being excellent in the water-impermeability can be obtained relatively economically. (N.H.)

33 CFR 151.2035 - What are the required ballast water management practices for my vessel?

Science.gov (United States)

2010-07-01

... water management practices for my vessel? 151.2035 Section 151.2035 Navigation and Navigable Waters...) must employ at least one of the following ballast water management practices: (1) Perform complete... SUBSTANCES, GARBAGE, MUNICIPAL OR COMMERCIAL WASTE, AND BALLAST WATER Ballast Water Management for Control of...

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

Science.gov (United States)

2010-07-01

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

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

Science.gov (United States)

2010-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

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

International Nuclear Information System (INIS)

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

1996-10-01

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

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

International Nuclear Information System (INIS)

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

1996-10-01

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

Sheet production apparatus for removing a crystalline sheet from the surface of a melt using gas jets located above and below the crystalline sheet

Energy Technology Data Exchange (ETDEWEB)

Kellerman, Peter L.; Thronson, Gregory D.

2017-06-14

In one embodiment, a sheet production apparatus comprises a vessel configured to hold a melt of a material. A cooling plate is disposed proximate the melt and is configured to form a sheet of the material on the melt. A first gas jet is configured to direct a gas toward an edge of the vessel. A sheet of a material is translated horizontally on a surface of the melt and the sheet is removed from the melt. The first gas jet may be directed at the meniscus and may stabilize this meniscus or increase local pressure within the meniscus.

Radioactive liquid containing vessel

International Nuclear Information System (INIS)

Sakurada, Tetsuo; Kawamura, Hironobu.

1993-01-01

Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

Science.gov (United States)

Yamaji, Akifumi; Li, Xin

2016-08-01

Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

Solid-state track recorder neutron dosimetry in light water reactor pressure vessel surveillance mockups

International Nuclear Information System (INIS)

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

1984-09-01

Solid-State Track Recorder (SSTR) measurements of neutron-induced fission rates have been made in several pressure vessel mockup facilities as part of the US Nuclear Regulatory Commission's (NRC) Light Water Reactor Pressure Vessel Surveillance Dosimetry Improvement Program (LWR-PV-SDIP). The results of extensive physics-dosimetry measurements made at the Pool Critical Assembly (PCA) at Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN are summarized. Included are 235 U, 238 U, 237 Np and 232 Th fission rates in the PCA 12/13, 8/7, and 4/12 SSC configurations. Additional low power measurements have been made in an engineering mockup at the VENUS critical assembly at CEN-SCK, Mol, Belgium. 237 Np and 238 U fission rates were made at selected locations in the VENUS mockup, which models the in-core and near-core regions of a pressurized water reactor (PWR). Absolute core power measurements were made at VENUS by exposing solid-state track recorders (SSTRs) to polished fuel pellets within in-core fuel pins. 8 references, 4 figures, 10 tables

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.

Analysis of three ex-vessel loss-of-coolant accidents in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-01-01

An ex-vessel LOCA may be caused by a rupture of a cooling pipe located outside the vacuum vessel. No plasma shutdown and no other counteractions have been assumed in order to study the worst case conditions of the accidents. The next three ex-vessel LOCAs in the primary cooling system of the first wall have been analysed: 1. a large break ex-vessel LOCA caused by a rupture of the cold leg (inner diameter 0.314 m) of the main circuit; 2. an intermediate break ex-vessel LOCA caused by a rupture of a sector inlet feeder (inner diameter 0.158 m); 3. an intermediate break ex-vessel LOCA caused by a rupture of the surge line (inner diameter 0.180 m) of the pressurizer. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the first two scenarios, melting in the first wall starts about 90 s after break initiation. In the third scenario, melting in the first wall start about 323 s after break initiation. Special emphasis has been paid to the characteristics of the break flows, the transient thermal-hydraulic behaviour of the cooling system, and the temperature development in the first wall. (orig.)

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.

Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

International Nuclear Information System (INIS)

Vlasichev, G.N.

1994-01-01

Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs

Removing vessels from the water for biofouling treatment has the potential to introduce mobile non-indigenous marine species.

Science.gov (United States)

Coutts, Ashley D M; Valentine, Joseph P; Edgar, Graham J; Davey, Adam; Burgess-Wilson, Bella

2010-09-01

Vessels found contaminated with biofouling non-indigenous marine species are predominantly removed from the water and treated in vessel maintenance facilities (i.e., slipways, travel lifts and dry-docks). Using pre-fouled settlement plates to simulate a vessel's removal from the water for treatment, we demonstrate that a range of mobile organisms (including non-indigenous marine species) may be lost to the marine environment as a consequence of this process. We also determined that different levels of biofouling (primary, secondary and tertiary) and emersion durations (0.5, 5 and 15 min) affected the abundance and composition of mobile taxa lost to the marine environment. Primary biofouling plates lost 3.2% of total animals, secondary plates lost 19.8% and tertiary plates lost 8.2%, while hanging duration had only minor effects. The results suggest that removing vessels contaminated with biofouling non-indigenous marine species from the water for treatment may not be as biosecure as is currently recognised. Copyright 2010 Elsevier Ltd. All rights reserved.

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.

A study on ex-vessel steam explosion for a flooded reactor cavity of reactor scale - 15216

International Nuclear Information System (INIS)

Song, S.; Yoon, E.; Kim, Y.; Cho, Y.

2015-01-01

A steam explosion can occur when a molten corium is mixed with a coolant, more volatile liquid. In severe accidents, corium can come into contact with coolant either when it flows to the bottom of the reactor vessel and encounters the reactor coolant, or when it breaches the reactor vessel and flows into the reactor containment. A steam explosion could then threaten the containment structures, such as the reactor vessel or the concrete walls/penetrations of the containment building. This study is to understand the shortcomings of the existing analysis code (TEXAS-V) and to estimate the steam explosion loads on reactor scale and assess the effect of variables, then we compared results and physical phenomena. Sensitivity study of major parameters for initial condition is performed. Variables related to melt corium such as corium temperature, falling velocity and diameter of melt are more important to the ex-vessel steam explosion load and the steam explosion loads are proportional to these variables related to melt corium. Coolant temperature on reactor cavity has a specific area to increase the steam explosion loads. These results will be used to evaluate the steam explosion loads using ROAAM (Risk Oriented Accident Analysis Methodology) and to develop the evaluation methodology of ex-vessel steam explosion. (authors)

Pressure vessel for a BWR type reactor

International Nuclear Information System (INIS)

Shimamoto, Yoshiharu.

1980-01-01

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

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