Method of reducing the hazard which may occur as a consequence of a reactor core meltdown

International Nuclear Information System (INIS)

Donne, M.D.; Dorner, S.; Schumacher, G.

1985-01-01

The core melt resulting from a meltdown accident of a GFB, LWR or LMFRR is collected by a core catcher from graphite placed below the core. The core melt is penetrating step by step into a borate store in the collecting vessel and is dissolving in it. Therefore the borate at the same time will absorb the decay heat. In order to remove the solidified and cooled down melted mass water is applied eliminating the borate. The remaining oxide states of the powdery core is sucked off again from the core catcher together with the water. The borate store (e.g. alkali borate) itself consists of separate layers with shaped parts, the coverings of which are made of steel, iron, cast iron, nickel, iron or nickel alloys, ceramic material or glass. (orig./PW)

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

International Nuclear Information System (INIS)

Ginsberg, T.; Chen, J.C.

1984-01-01

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

Concepts for passive heat removal and filtration systems under core meltdown conditions

International Nuclear Information System (INIS)

Wilhelm, J.G.; Neitzel, H.-J.

1993-01-01

The objective of the new containment concept being developed by KfK is the complete passive enclosure of a power reactor after a core meltdown accident by means of a solid containment structure and passive removal of the decay heat. This is to be accomplished by cooling the containment walls with ambient air, with thermoconvection as the driving force. The concept of the containment is described. Data are given of the heat removal and the requirements for filtration of the exhaust air, which is contaminated due to the leak rate assumed for the inner containment. The concept for the filter system is described. Various solutions for reduction of the large volumetric flow to be filtered are discussed. 3 refs., 8 figs

On the failure modes of alternative containment designs following postulated core meltdown

International Nuclear Information System (INIS)

Chan, C.K.; Knee, H.E.; Okrent, D.

1977-01-01

The containment response to a postulated core meltdown accident in a PWR ice condenser containment, a BWR Mark III containment and a BWR non-inerted Mark I containment has been examined to see if the WASH-1400 containment failure mode judgement for the Surry large, dry containment and the Peach Bottom Mark I inerted-containment are likely to be appropriate for these alternative containment plant designs. For the PWR, the representative accident chosen for the analysis is a large cold leg break accompanied by a loss of all electric power while the BWR respresentative event chosen is a recirculation line break without adequate core cooling function. Two containment event paths are studied for each of these two cases, depending on whether or not containment vapor suppression function is assumed to be available. Both the core and the containment pressure and temperature response to the accident events are computed for the four time intervals which characterize (a) blowdown of the pipe break, (b) core melt, (c) vessel melt-through, and (d) containment foundation penetration. The calculations are based on a best esimate of the most probable sequence, but certain phenomena and events were followed down multiple tracks. It appears that the non-inerted Mark I containment is not so vulnerable to overpressurization from hydrogen burning as the Mark III; however, acceptable temperatures may be exceeded. (Auth.)

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

International Nuclear Information System (INIS)

Ginsberg, T.; Chen, J.C.

1984-01-01

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

How did Fukushima-Dai-ichi core meltdown change the probability of nuclear accidents?

International Nuclear Information System (INIS)

Escobar Rangel, Lina; Leveque, Francois

2012-10-01

How to predict the probability of a nuclear accident using past observations? What increase in probability the Fukushima Dai-ichi event does entail? Many models and approaches can be used to answer these questions. Poisson regression as well as Bayesian updating are good candidates. However, they fail to address these issues properly because the independence assumption in which they are based on is violated. We propose a Poisson Exponentially Weighted Moving Average (PEWMA) based in a state-space time series approach to overcome this critical drawback. We find an increase in the risk of a core meltdown accident for the next year in the world by a factor of ten owing to the new major accident that took place in Japan in 2011. (authors)

Evaluation of containment failure modes and fission product releases during core meltdown accidents in a BWR with a Mark III containment

International Nuclear Information System (INIS)

Ludewig, H.; Yu, W.S.; Jaung, R.; Pratt, W.T.

1985-01-01

An assessment is described of potential failure modes and fission product releases for a large number of postulated core meltdown accidents in a BWR with a Mark III containment. For this containment design, the most important failure mode was found to be due to hydrogen related phenomena. A one-dimensional lumped parameter computer code has been developed and used to determine the probability of various hydrogen phenomena for a range of postulated core meltdown sequences. Potential containment loads have been estimated and compared against the containment capacity to determine the probability of containment failure. The fission product release assessment began by using the MARCH/CORRAL system of codes with key input parameters varied over a reasonable range. The parameters relate to primary system retention, re-emission, pool scrubbing, and fission product release in-vessel vs ex-vessel. The final step used more mechanistic calculations based on the system of codes recently developed under sponsorship of the Accident Source Term Program Office, NRC, and compares these predictions with the range of releases calculated in the sensitivity study

Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

International Nuclear Information System (INIS)

Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

1985-01-01

A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

Analyses of containment loading by hydrogen burning during hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Bracht, K.; Tiltmann, M.

1983-01-01

The possibility of occurance of violent hydrogen burning during a LWR meltdown accident and its consequences to containment atmosphere conditions are discussed. Two accident sequences with low and high system pressure during the in-vessel-melt phase of a meltdown accident are considered. In both sequences only deflagration, but no detonation may become possible, presuming homogeneity of the containment atmospheres. In a low pressure szenario the pressure increase due to deflagration will not reach the failure pressure of the containment, if combustion takes place when the flammability limit is reached. For the special situation of a rapid release of steam and hydrogen after a high-pressure failure of a reactor pressure vessel, calculations with a multicompartment code show that the possibility for hydrogen burning does not exist. Thus, an additional augmentation of the steam spike as a consequence of the failure of the pressure vessel cannot occur. (orig.)

On-site releases of noble gases and iodine in the event of core meltdown in a swimming pool reactor

International Nuclear Information System (INIS)

Montaignac, E. de.

1976-10-01

Research aimed at defining a standard model accident for swimming pool type reactors, has led to the adoption to the so-called BORAX accident which involves complete meltdown of the reactor core. This type of accident-an accident related to dimensional problems- is useful for calculations concerning reactor components which have to withstand the mechanical forces resulting from the accident. A study of the radiobiological consequences of this type of accident, involving the entire reactor core, required research to determine as accurately as possible how the iodine, noble gases and solid fission products are distributed between the melted core and the site. The joint document in the annexure served as the basis for discussion at the meeting (BEVS/SESR) on 9th March 1973, at which the SESR set the standard parameter values to be used for estimating fission product distributions on the site. (author)

Transport of nuclides during a core meltdown accident, with consideration of filtered venting

International Nuclear Information System (INIS)

Haeggblom, H.

1981-01-01

A BWR core meltdown accident has been studied with respect to the transport of radioactive and nonactive gases and aerosols. A system consisting of a containment with an outer stone condenser in three parts was considered. Calculations of the aerosol behaviour have been made with the computer programme NAUA and HAARM-3, assuming one single compartment. Results from these calculations have been used for multicompartment calculations with CORRAL II. The code was modified so that particles of different sizes could be considered in the different compartments, and the time dependence of the particles can be arbitrary. In addition to the aerosol transport and deposition, the corresponding quantities for elemental iodine were calculated. It was concluded, that if the total volume of the condenser system is of the order of 10 5 m 3 , practically all elemental iodine and particles can be retained in the system. The only leakage to the environment will be caused by inefficient sealing during the first five hours. The pressure can never damage the condenser. (author)

Aerosol retention during SGTR meltdown sequences under different conditions of tubes vibration

International Nuclear Information System (INIS)

Delgado Tardaguila, R.; Herranz, L. E.

2014-01-01

The containment bypass of the fission products during SGTR meltdown sequences makes this scenario be a significant risk contributor in PWRs. The EU-SGTR, ARTIST 1 and 2 and the on-going ARTIST-extension programs have investigated the potential attenuation of the source term reaching the secondary side of a failed SG even in the absence of water. (Author)

Core catcher for nuclear reactor core meltdown containment

International Nuclear Information System (INIS)

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

1978-01-01

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

Investigation on multilayer failure mechanism of RPV with a high temperature gradient from core meltdown scenario

Energy Technology Data Exchange (ETDEWEB)

Jianfeng, Mao, E-mail: jianfeng-mao@163.com [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China); Xiangqing, Li [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Shiyi, Bao, E-mail: bsy@zjut.edu.cn [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China); Lijia, Luo [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Zengliang, Gao [Institute of Process Equipment and Control Engineering, Zhejiang University of Technology, Hangzhou, Zhejiang 310032 (China); Engineering Research Center of Process Equipment and Remanufacturing, Ministry of Education (China)

2016-12-15

Highlights: • The multilayer failure mechanism is investigated for RPV under CHF. • Failure time and location of RPV are predicted under various SA scenarios. • The structural behaviors are analyzed in depth for creep and plasticity. • The effect of internal pressure and temperature gradient is considered. • The structural integrity of RPV is secured within the required 72 creep hours. - Abstract: The Fukushima accident shows that in-vessel retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still exists inside the reactor pressure vessel (RPV). In the traditional concept of IVR, the pressure is supposed to successfully be released, and the temperature distributed among the wall thickness is assumed to be uniform. However, this concept is seriously challenged by reality of Fukushima accident with regard to the existence of both internal pressure and high temperature gradient. Therefore, in order to make the IVR mitigation strategy succeed, the numerical investigation of the lower head behavior and its failure has been performed for several internal pressures under high temperature gradient. According to some requirements in severe accident (SA) management of RPV, it should be ensured that the IVR mitigation takes effect in preventing the failure of the structure within a period of 72 h. Subsequently, the failure time and location have to be predicted under the critical heat flux (CHF) loading condition for lower head, since the CHF is limit thermal boundary before the melt-through of RPV. In illustrating the so called ‘multilayer failure mechanism’, the structural behaviors of RPV are analyzed in terms of the stress, creep strain, deformation, damage on selected paths.

Investigation on multilayer failure mechanism of RPV with a high temperature gradient from core meltdown scenario

International Nuclear Information System (INIS)

Jianfeng, Mao; Xiangqing, Li; Shiyi, Bao; Lijia, Luo; Zengliang, Gao

2016-01-01

Highlights: • The multilayer failure mechanism is investigated for RPV under CHF. • Failure time and location of RPV are predicted under various SA scenarios. • The structural behaviors are analyzed in depth for creep and plasticity. • The effect of internal pressure and temperature gradient is considered. • The structural integrity of RPV is secured within the required 72 creep hours. - Abstract: The Fukushima accident shows that in-vessel retention (IVR) of molten core debris has not been appropriately assessed, and a certain pressure (up to 8.0 MPa) still exists inside the reactor pressure vessel (RPV). In the traditional concept of IVR, the pressure is supposed to successfully be released, and the temperature distributed among the wall thickness is assumed to be uniform. However, this concept is seriously challenged by reality of Fukushima accident with regard to the existence of both internal pressure and high temperature gradient. Therefore, in order to make the IVR mitigation strategy succeed, the numerical investigation of the lower head behavior and its failure has been performed for several internal pressures under high temperature gradient. According to some requirements in severe accident (SA) management of RPV, it should be ensured that the IVR mitigation takes effect in preventing the failure of the structure within a period of 72 h. Subsequently, the failure time and location have to be predicted under the critical heat flux (CHF) loading condition for lower head, since the CHF is limit thermal boundary before the melt-through of RPV. In illustrating the so called ‘multilayer failure mechanism’, the structural behaviors of RPV are analyzed in terms of the stress, creep strain, deformation, damage on selected paths.

Safety analysis of a stratified reactor foundation subject to core meltdown

International Nuclear Information System (INIS)

Laermann, K.H.; Berhalter, D.; Wieching, G.

1983-01-01

Comprehensive calculations were made to examine a foundation plate under static load with non-linear material behaviour. In particular, deformation and crack behaviour of the plate were observed. The calculations hereunder, made to calculate the time of rupture of the plate with the foundation exposed to high temperatures, with and without cavity formation, prove that high temperatures are not as relevant as the melt-down of the layers of armour on top of the middle of the plate. The consequences of this survey are: to protect the top armour of the plate centre and not to use it as support for any structural components. This free space could then be used to instal said multi-layer plate. (orig./HP) [de

Nuclear reactor core catcher

International Nuclear Information System (INIS)

1977-01-01

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

Recriticality, a Key Phenomenon to Investigate in Core Disruptive Accident Scenarios of Current and Future Fast Reactor Designs

International Nuclear Information System (INIS)

Maschek, W.; Rineiski, A.; Flad, M.; Kriventsev, V.; Gabrielli, F.; Morita, K.

2012-01-01

Final comments and conclusions: • Modern plants, should have performed better under Fukushima type event. • In future fast reactor systems significantly higher active and passive safety features are installed, which should cope with events like Fukushima. • One important lesson: put a focus on rare initiators, accident routes and consequences that are neither expected nor have been observed, events that are categorized under ‘black swans’. • Importance of severe accident research demonstrated - both analytically and experimentally for assessing and interpreting accident scenarios and developments. Precondition for developing preventive & mitigative safety measures. Passive safety measures are in the focus of advanced design options and must work under conditions of multiple loads and aggravating events. • Fast reactor systems behavior as the SFR under severe accident conditions: – In fast spectrum systems as the SFR the core is not in its neutronically most reactive configuration and SFRs may be loaded with MAs for waste management; – Recriticalities have a high probability because of the higher enrichment levels; – Short time scales have to be envisioned for core melt-down; – Decay heat levels might be significantly higher, if MA bearing fuel is involved. • Improve design by measures for prevention and/or mitigation of recriticalities; – High reliability of simulations required for proof; • Assessment of fuel relocated on peripheral structures; • Preventive/mitigating measures should not replace containment measures

Simulation experiments concerning core meltdown

International Nuclear Information System (INIS)

Werle, H.

1979-01-01

A gas stream causes a remarkable increase in the interfacial heat flux (by a factor of 8 for v = 0.63 cm/s, v = gas volume flux/horizontal area). The most important characteristics of the system investigated (silicon oil/wood metal) are relatively similar to those of a core melt, Therefore a remarkable increase of the interfacial heat transfer by the gas release may be expected also for a core melt, compared with earlier investigations at the system silicon oil/water the influence of a gas stream is nevertheless remarkably lower for silicon oil/wood metal. This shows that the density ratio plays an important role. (orig./RW) [de

Caesium-rich micro-particles: A window into the meltdown events at the Fukushima Daiichi Nuclear Power Plant

Science.gov (United States)

Furuki, Genki; Imoto, Junpei; Ochiai, Asumi; Yamasaki, Shinya; Nanba, Kenji; Ohnuki, Toshihiko; Grambow, Bernd; Ewing, Rodney C.; Utsunomiya, Satoshi

2017-02-01

The nuclear disaster at the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011 caused partial meltdowns of three reactors. During the meltdowns, a type of condensed particle, a caesium-rich micro-particle (CsMP), formed inside the reactors via unknown processes. Here we report the chemical and physical processes of CsMP formation inside the reactors during the meltdowns based on atomic-resolution electron microscopy of CsMPs discovered near the FDNPP. All of the CsMPs (with sizes of 2.0-3.4 μm) comprise SiO2 glass matrices and ~10-nm-sized Zn-Fe-oxide nanoparticles associated with a wide range of Cs concentrations (1.1-19 wt% Cs as Cs2O). Trace amounts of U are also associated with the Zn-Fe oxides. The nano-texture in the CsMPs records multiple reaction-process steps during meltdown in the severe FDNPP accident: Melted fuel (molten core)-concrete interactions (MCCIs), incorporating various airborne fission product nanoparticles, including CsOH and CsCl, proceeded via SiO2 condensation over aggregates of Zn-Fe oxide nanoparticles originating from the failure of the reactor pressure vessels. Still, CsMPs provide a mechanism by which volatile and low-volatility radionuclides such as U can reach the environment and should be considered in the migration model of Cs and radionuclides in the current environment surrounding the FDNPP.

ASTRID core: Design objectives, design approach, and R&D in support

International Nuclear Information System (INIS)

Mignot, G.; Devictor, N.

2012-01-01

ASTRID core design is mainly guided by safety objectives: 1. Prevention of the core meltdown accident: To prevent meltdown accidents: - by a natural behavior of the core and the reactor (no actuation of the two shutdown systems); - with adding passive complementary systems if natural behavior is not sufficient for some transient cases. 2. Mitigation of the fusion accident: To garantee that core fusion accidents don’t lead to significant mechanical energy release, whatever initiator event: - by a natural core behavior; - with adding specific mitigation dispositions in case of natural behavior is not suffficient

Reaction- and melting behaviour of LWR-core components UO2, Zircaloy and steel during the meltdown period

International Nuclear Information System (INIS)

Hofmann, P.

1976-07-01

The reaction behaviour of the UO 2 , Zircaloy-4 and austenitic steel core components was investigated as a function of temperature (till melting temperatures) under inert and oxidizing conditions. Component concentrations varied between that of Corium-A (65 wt.% UO 2 , 18% Zry, 17% steel) and that of Corium-E (35 wt.% UO 2 , 10% Zry, 55% steel). In addition, Zircaloy and stainless steel were used with different degrees of oxidation. The paper describes systematically the phases that arise during heating and melting. The integral composition of the melts and the qualitative as well as quantitative analysis of the phases present in solidified corium are given. In some cases melting points have been determined. The reaction and melting behaviour of the corium specimens strongly depends on the concentration and on the degree of oxidation of the core components. First liquid phases are formed at the Zry-steel interface at about 1,350 0 C. The maximum temperatures of about 2,500 0 C for the complete melting of the corium-specimens are well below the UO 2 melting point. Depending on the steel content and/or degree of oxidation of Zry and steel, a homogeneous metallic or oxide melt or two immiscible melts - one oxide and the other metallic - are obtained. During the melting experiments performed under inert gas conditions the chemical composition of the molten specimens generally change by evaporation losses of single elements, especially of uranium, zirconium and oxygen. The total weight losses go up to 30%; under oxidizing conditions they are substantially smaller due to the occurrence of different phases. In air or water vapor, the occurrence of the phases and the melting behaviour of the core components are strongly influenced by the oxidation rate and the oxygen supply to the surface of the melt. In the case of the hypothetical core melting accident, a heterogeneous melt (oxide and metallic) is probable after the meltdown period. (orig./RW) [de

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

Eigen's Error Threshold and Mutational Meltdown in a Quasispecies Model

OpenAIRE

Bagnoli, F.; Bezzi, M.

1998-01-01

We introduce a toy model for interacting populations connected by mutations and limited by a shared resource. We study the presence of Eigen's error threshold and mutational meltdown. The phase diagram of the system shows that the extinction of the whole population due to mutational meltdown can occur well before an eventual error threshold transition.

Investigation of the different scenarios occurring in a PWR in case of a TMLB accident

International Nuclear Information System (INIS)

Pochard, R.; Dufresne, J.; Autrusson, B.

1988-10-01

Severe accidents in light water reactors fall into one of two main categories, depending on whether or not core meltdown is accompanied by a pressure buildup in the primary system. The way in which the accident develops is, in fact, largely conditioned by this pressure aspect: temperature distribution in the core and primary system resulting from natural convection gas streams; fuel clad failure mode, etc... One major effect of pressure buildup on the accident scenario is primary system failure under the combined actions of pressure and temperature. The purpose of the present paper is to present, after a detailed thermalhydraulic study, an analysis of the timing and location of the system failures in case of a TMLB accident on CPY french type reactor

Oxidation during reflood of reactor core with melting cladding

Energy Technology Data Exchange (ETDEWEB)

Siefken, L.J.; Allison, C.M.; Davis, K.L. [and others

1995-09-01

Models were recently developed and incorporated into the SCDAP/RELAP5 code for calculating the oxidation of fuel rods during cladding meltdown and reflood. Experiments have shown that a period of intense oxidation may occur when a hot partially oxidized reactor core is reflooded. This paper offers an explanation of the cladding meltdown and oxidation processes that cause this intense period of oxidation. Models for the cladding meltdown and oxidation processes are developed. The models are assessed by simulating a severe fuel damage experiment that involved reflood. The models for cladding meltdown and oxidation were found to improve calculation of the temperature and oxidation of fuel rods during the period in which hot fuel rods are reflooded.

Mutational meltdown in laboratory yeast populations

NARCIS (Netherlands)

Zeyl, C.; Mizesko, M.; Visser, de J.A.G.M.

2001-01-01

In small or repeatedly bottlenecked populations, mutations are expected to accumulate by genetic drift, causing fitness declines. In mutational meltdown models, such fitness declines further reduce population size, thus accelerating additional mutation accumulation and leading to extinction. Because

Thermohydraulics in a high-temperature gas-cooled reactor primary loop during early phases of unrestricted core-heatup accidents

International Nuclear Information System (INIS)

Kroeger, P.G.; Colman, J.; Hsu, C.J.

1983-01-01

In High Temperature Gas Cooled Reactor (HTGR) siting considerations, the Unrestricted Core Heatup Accidents (UCHA) are considered as accidents of highest consequence, corresponding to core meltdown accidents in light water reactors. Initiation of such accidents can be, for instance, due to station blackout, resulting in scram and loss of all main loop forced circulation, with none of the core auxiliary cooling system loops being started. The result is a slow but continuing core heatup, extending over days. During the initial phases of such UCHA scenarios, the primary loop remains pressurized, with the system pressure slowly increasing until the relief valve setpoint is reached. The major objectives of the work described here were to determine times to depressurization as well as approximate loop component temperatures up to depressurization

Study of diluting and absorber materials to control the reactivity during a postulated core meltdown accident in generation IV reactors

International Nuclear Information System (INIS)

Plevacova, Kamila

2010-01-01

In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic points of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author)

GLOBAL ECONOMIC MELTDOWN AND THE NIGERIAN ...

African Journals Online (AJOL)

global financial meltdown, marked by the collapse of hitherto revered and great financial ... Indeed it is an integral part of a market (capitalist) economy. .... competitively in economic production, so also is the deficit in its trade with the rest of the.

Synthesis of the IRSN report on the topic of water way answers to implement in case of accident with core meltdown occurring on operating pressurized water nuclear reactors

International Nuclear Information System (INIS)

2009-06-01

This report briefly discusses the efficiency of technical measures adopted for the implementation of water ways as answers to an accident with core meltdown in operating pressurized water nuclear reactors. While mentioning the importance of the hydro-geological characteristics of the various sites, the IRSN asks EDF to plan and implement means to prevent any rejection through water ways for some of these sites, to investigate the possibility of building a geotechnical enclosure, to define a storing-control-treatment-rejection chain which would guarantee an efficient management of the water to be pumped, to study retention phenomena for strontium and caesium isotopes in sands and gravels

Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

International Nuclear Information System (INIS)

Rakhi; Sharma, Anil Kumar; Velusamy, K.

2017-01-01

Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

Laboratory studies of the meltfront propagation in a borax core-catcher

International Nuclear Information System (INIS)

Dalle Donne, M.; Werle, H.

1980-08-01

A series of seven laboratory experiments concerning the meltdown of a borax core catcher have been performed. By the selection of the simulant materials the most important thermophysical properties of the core catcher materials were taken into account. Fission product heating of the molten core masses was simulated by electrolytically heating of the molten region. The experiments reveal interesting details of the phenomena to be expected during melt-down of a borax core catcher, especially on the flow pattern, the mixing processes of molten materials and the layer formation the melt. The most interesting result is that the ratio of downward to sideward melting rate is heavily reduced by high melting barriers and that a cubic structure of barriers will not equalize downward and sideward melting rates. A super 8 film is available as additional information. (orig.) [de

Destruction of the BETA experimental facility for core meltdown experiments in the Karlsruhe Nuclear Research Center on 21 March 1992

International Nuclear Information System (INIS)

Feige, K.D.

1992-01-01

The BETA experiment V 6.2 was intended to yield information on the processes involved in a lateral containment meltdown starting in a concrete wall with external water cooling. The unexpected overpressure that caused the explosion occurred 1896 seconds after the melt had been fed into the crucible, inducing the melt-water interaction. The explosion destroyed only the inner space of the facility. (orig.) [de

Risk reduction of core-melt accidents in advaned CAPRA burner cores

International Nuclear Information System (INIS)

Maschek, W.; Struwe, D.; Eigemann, M.

1997-01-01

As part of the CAPRA Program (Consommation Accrue de Plutonium dans les RApides) the feasibility of fast reactors is investigated to burn plutonium and also to destruct minor actinides. The design of CAPRA cores shows significant differences compared to conventional cores. Especially the high Pu-enrichment has an important influence on the core melt-down behavior and the associated recriticality risk. To cope with this risk, inherent design features and special measures/devices are investigated for their potential of early fuel discharge to reduce the criticality of the reactor core. An assessment of such measures/devices is given and experimental needs are formulated. 11 refs., 5 figs

Assessment of uncertainties in core melt phenomenology and their impact on risk at the Z/IP facilities

International Nuclear Information System (INIS)

Pratt, W.T.; Ludewig, H.; Bari, R.A.; Meyer, J.F.

1983-01-01

An evaluation of core meltdown accidents in the Z/IP facilities has been performed. Containment event trees have been developed to relate the progression of a given accident to various potential containment building failure modes. An extensive uncertainty analysis related to core melt phenomenology has been performed. A major conclusion of the study is that large variations in parameters associated with major phenomenological uncertainties have a relatively minor impact on risk when external initiators are considered. This is due to the inherent capability fo the Z/IP containment buildings to contain a wide range of core meltdown accidents. 12 references, 2 tables

How to arrest a core meltdown accident (doing nothing)

International Nuclear Information System (INIS)

Baron, Jorge H.

2000-01-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)

Assessment of accident energetics in LMFBR core-disruptive accidents

International Nuclear Information System (INIS)

Fauske, H.K.

1977-01-01

An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggests that such events are highly unlikely following a postulated core meltdown event

Lack of Critical Slowing Down Suggests that Financial Meltdowns Are Not Critical Transitions, yet Rising Variability Could Signal Systemic Risk

Science.gov (United States)

Hoarau, Quentin

2016-01-01

Complex systems inspired analysis suggests a hypothesis that financial meltdowns are abrupt critical transitions that occur when the system reaches a tipping point. Theoretical and empirical studies on climatic and ecological dynamical systems have shown that approach to tipping points is preceded by a generic phenomenon called critical slowing down, i.e. an increasingly slow response of the system to perturbations. Therefore, it has been suggested that critical slowing down may be used as an early warning signal of imminent critical transitions. Whether financial markets exhibit critical slowing down prior to meltdowns remains unclear. Here, our analysis reveals that three major US (Dow Jones Index, S&P 500 and NASDAQ) and two European markets (DAX and FTSE) did not exhibit critical slowing down prior to major financial crashes over the last century. However, all markets showed strong trends of rising variability, quantified by time series variance and spectral function at low frequencies, prior to crashes. These results suggest that financial crashes are not critical transitions that occur in the vicinity of a tipping point. Using a simple model, we argue that financial crashes are likely to be stochastic transitions which can occur even when the system is far away from the tipping point. Specifically, we show that a gradually increasing strength of stochastic perturbations may have caused to abrupt transitions in the financial markets. Broadly, our results highlight the importance of stochastically driven abrupt transitions in real world scenarios. Our study offers rising variability as a precursor of financial meltdowns albeit with a limitation that they may signal false alarms. PMID:26761792

Evaluation of Effect of Global Economic Meltdown on Capital Market Performance

Directory of Open Access Journals (Sweden)

ONAOLAPO ADEKUNLE RAHMAN

2013-07-01

Full Text Available The recent financial crisis that loomed the global economy was considered more inclusive than any other period of financial turmoil in the past 60 years. This paper evaluates the implications of the global economic meltdown on the Nigerian Capital Market Performance using the market capitalization of the Nigerian Stock Exchange as a major indicator. When the global economic meltdown came, it poses a recession on the Market Capitalization and the volume of share index of the Nation. This study depends entirely on secondary data in form of annual aggregate time series data of Market capitalization (dependent variable, exchange rate, interest rate, inflation rate, market share index with Dummy variable to represent the period of economic crisis. Ordinary least square of multiple regressions was used to analyze the data into econometric model while F-statistics was used to test for the formulated hypothesis. This study depicts that the global economic meltdown has a negative effect on the Capital Market Performance. It was therefore recommended that the Federal government and the regulatory agencies (CBN, NSE, SEC etc. should come up with intervention and fiscal policies that will suppress these effects and jumpstart the capital market and that the policies should be properly implemented and monitored.

CLASS: Core Library for Advanced Scenario Simulations

International Nuclear Information System (INIS)

Mouginot, B.; Thiolliere, N.

2015-01-01

The nuclear reactor simulation community has to perform complex electronuclear scenario simulations. To avoid constraints coming from the existing powerful scenario software such as COSI, VISION or FAMILY, the open source Core Library for Advanced Scenario Simulation (CLASS) has been developed. The main asset of CLASS is its ability to include any type of reactor, whether the system is innovative or standard. A reactor is fully described by its evolution database which should contain a set of different validated fuel compositions in order to simulate transitional scenarios. CLASS aims to be a useful tool to study scenarios involving Generation-IV reactors as well as innovative fuel cycles, like the thorium cycle. In addition to all standard key objects required by an electronuclear scenario simulation (the isotopic vector, the reactor, the fuel storage and the fabrication units), CLASS also integrates two new specific modules: fresh fuel evolution and recycled fuel fabrication. The first module, dealing with fresh fuel evolution, is implemented in CLASS by solving Bateman equations built from a database induced cross-sections. The second module, which incorporates the fabrication of recycled fuel to CLASS, can be defined by user priorities and/or algorithms. By default, it uses a linear Pu equivalent-method, which allows predicting, from the isotopic composition, the maximum burn-up accessible for a set type of fuel. This paper presents the basis of the CLASS scenario, the fuel method applied to a MOX fuel and an evolution module benchmark based on the French electronuclear fleet from 1977 to 2012. Results of the CLASS calculation were compared with the inventory made and published by the ANDRA organisation in 2012. For UOX used fuels, the ANDRA reported 12006 tonnes of heavy metal in stock, including cooling, versus 18500 tonnes of heavy metal predicted by CLASS. The large difference is easily explained by the presence of 56 tonnes of plutonium already separated

Proposal for computer investigation of LMFBR core meltdown accidents

International Nuclear Information System (INIS)

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

1974-01-01

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

Apparatus for controlling nuclear core debris

Science.gov (United States)

Jones, Robert D.

1978-01-01

Nuclear reactor apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling.

Apparatus for controlling nuclear core debris

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

Disclosed is an apparatus for containing, cooling, and dispersing reactor debris assumed to flow from the core area in the unlikely event of an accident causing core meltdown. The apparatus includes a plurality of horizontally disposed vertically spaced plates, having depressions to contain debris in controlled amounts, and a plurality of holes therein which provide natural circulation cooling and a path for debris to continue flowing downward to the plate beneath. The uppermost plates may also include generally vertical sections which form annular-like flow areas which assist the natural circulation cooling

Apparatus for controlling molten core debris

International Nuclear Information System (INIS)

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

1972-01-01

Disclosed is an apparatus for containing, cooling, diluting, dispersing and maintaining subcritical the molten core debris assumed to melt through the bottom of a nuclear reactor pressure vessel in the unlikely event of a core meltdown. The apparatus is basically a sacrificial bed system which includes an inverted conical funnel, a core debris receptacle including a spherical dome, a spherically layered bed of primarily magnesia bricks, a cooling system of zig-zag piping in graphite blocks about and below the bed and a cylindrical liner surrounding the graphite blocks including a steel shell surrounded by firebrick. Tantalum absorber rods are used in the receptacle and bed. 9 claims, 22 figures

Surface analytical investigations of the release behaviour of volatile fission products during simulated core meltdown accidents and of the reaction behaviour of iodine with silver surfaces

International Nuclear Information System (INIS)

Moers, H.

1986-07-01

The report presents the results of the analysis of aerosol particles formed in simulated laboratory scale core meltdown experiments. In addition the interaction of silver surfaces with gaseous molecular iodine and with iodide and molecular iodine in aqueous solution was investigated. The composition of the aerosol samples and the progress of the reactions mentioned were determined by use of surface analytical techniques (X-ray photoelectron spectroscopy, Auger electron spectroscopy, secondary ion mass spectroscopy). The major information can be evaluated from X-ray photoelectron spectra which exhibit chemical shifts of the photoelectron lines which allowing a discrimination between different chemical species of the same element. The analyses showed that iodine is present in the aerosol particles mainly as caesium iodide and, to a smaller fraction, as silver iodide. During the adsorption of gaseous molecular iodine at metallic silver surfaces a closed silver iodide overlayer is formed. In aqueous iodide solutions one observes chemisorption of the iodide anions up to a coverage of the metallic silver surface of about half a monolayer. Molecular iodine in aqueous solution is completely converted to silver iodide which covers the substrate irregularly. (orig./HP) [de

Lessons from the Russian meltdown : the economics of soft legal constraints

NARCIS (Netherlands)

Perotti, E.C.

2001-01-01

On August 17, 1998, Russia defaulted on its domestic public debt, declared a moratorium on the private banks foreign liabilities which was equivalent to an outright default, and abandoned its exchange rate regime. The depth of the Russian meltdown shocked the international markets, and precipitated

Lessons from the Russian meltdown : the economics of soft legal constraints

NARCIS (Netherlands)

Perotti, E.C.

2002-01-01

On August 17, 1998, Russia defaulted on its domestic public debt, declared a moratorium on the private banks foreign liabilities which was equivalent to an outright default, and abandoned its exchange rate regime. The depth of the Russian meltdown shocked the international markets, and precipitated

A Primer on Financial System Meltdown. The Economists' View

OpenAIRE

Hahn, Franz R.

2008-01-01

Ideologues are quick to explain the current financial meltdown: it's the markets, stupid. Economists agree but add: it's politics too, stupid. Ideologues agree but counter: first and foremost it's capitalism, stupid. Economists agree but reply: §$%&?!, stupid. This is where this short paper takes us: it makes an attempt to give a brief overview of the economists' views on the ongoing financial system crisis explaining "§$%&?!, stupid".

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

Lessons from the Russian meltdown : the economics of soft legal constraints

NARCIS (Netherlands)

Perotti, E.C.

2002-01-01

On 17 August 1998 Russia abandoned its exchange rate regime, defaulted on its domestic public debt and declared a moratorium on banks' foreign liabilities. This was equivalent to an outright default. The depth and speed of the Russian meltdown shocked the international markets and precipitated a

Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-01-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

Nuclear reactor core safety device

International Nuclear Information System (INIS)

Colgate, S.A.

1977-01-01

The danger of a steam explosion from a nuclear reactor core melt-down can be greatly reduced by adding a gasifying agent to the fuel that releases a large amount of gas at a predetermined pre-melt-down temperature that ruptures the bottom end of the fuel rod and blows the finely divided fuel into a residual coolant bath at the bottom of the reactor. This residual bath should be equipped with a secondary cooling loop

Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

Energy Technology Data Exchange (ETDEWEB)

Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

2016-04-15

Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

DEVELOPMENT AND DEPLOYMENT ASSESSMENT OF A MELT-DOWN PROOF MODULAR MICRO REACTOR (MDP-MMR)

Energy Technology Data Exchange (ETDEWEB)

Hawari, Ayman I.; Venneri, Francesco

2018-04-02

The objective of this project is to perform feasibility assessment and technology gap analysis and establish a development roadmap for an innovative and highly compact Micro Modular Reactor (MMR) concept that integrates power production, power conversion and electricity generation in a single unit. The MMR is envisioned to use fully ceramic micro-encapsulated (FCM) fuel, a particularly robust form of TRISO fuel, and to be gas-cooled (e.g., He or CO2) and capable of generating power in the range of 10 to 40 MW-thermal. It is designed to be absolutely melt-down proof (MDP) under all circumstances including complete loss of coolant scenarios with no possible release of radioactive material, to be factory produced, to have a cycle length of greater than 20 years, and to be highly proliferation resistant. In addition, it will be transportable, retrievable and suitable for use in remote areas. As such, the MDP-MMR will represent a versatile reactor concept that is suitable for use in various applications including electricity generation, process heat utilization and propulsion.

The acute effects of the thermogenic supplement Meltdown on energy expenditure, fat oxidation, and hemodynamic responses in young, healthy males

Directory of Open Access Journals (Sweden)

Cooke Matt

2008-12-01

Full Text Available Abstract The purpose of this study was to evaluate the effects of a thermogenic supplement, Meltdown, on energy expenditure, fat oxidation, and hemodynamics before and after maximal treadmill exercise. In a double-blind, randomized, placebo-controlled, cross-over design, 12 male participants underwent two testing sessions after consuming either the Meltdown or placebo supplement. While in a fasted state, participants rested for one hour, orally ingested either Meltdown or placebo and rested for another hour, performed a maximal treadmill exercise test, and then rested for another hour. Throughout the testing protocol, resting energy expenditure (REE and respiratory exchange ratio (RER were assessed. In addition, heart rate (HR and blood pressure (BP were assessed before and after exercise. Meltdown increased REE significantly more than placebo at 45 min (1.44 ± 0.25 vs. 1.28 ± 0.23 kcal/min; p = 0.003, 60 min (1.49 ± 0.28 vs. 1.30 ± 0.22 kcal/min; p = 0.025, and 120 min (1.51 ± 0.26 vs. 1.33 ± 0.27 kcals/min; p = 0.014 post-ingestion. Meltdown significantly decreased RER at 30 min (0.84 ± 0.03 vs. 0.91 ± 0.04; p = 0.022 and 45 min post-ingestion (0.82 ± 0.04 vs. 0.89 ± 0.05; p = 0.042, and immediately post-exercise (0.83 ± 0.05 vs. 0.90 ± 0.07; p = 0.009. Furthermore, over the course of the evaluation period, area under the curve assessment demonstrated that REE was significantly increased with Meltdown compared to placebo (992.5 ± 133.1 vs. 895.1 ± 296.1 kcals; p = 0.043, while RER was significantly less than placebo (5.55 ± 0.61 vs. 5.89 ± 0.44; p = 0.002 following ingestion. HR and BP were not significantly affected prior to exercise with either supplement (p > 0.05 and the exercise-induced increases for HR and BP decreased into recovery and were not different between supplements (p > 0.05. These data suggest that Meltdown enhances REE and fat oxidation more than placebo for several hours after ingestion in fully rested and

The risk of extinction - the mutational meltdown or the overpopulation

OpenAIRE

Malarz, K.

2006-01-01

The phase diagrams survival-extinction for the Penna model with parameters: (mutations rate)-(birth rate), (mutation rate)-(harmful mutations threshold), (harmful mutation threshold)-(minimal reproduction age) are presented. The extinction phase may be caused by either mutational meltdown or overpopulation. When the Verhulst factor is responsible for removing only newly born babies and does not act on adults the overpopulation is avoided and only genetic factors may lead to species extinction.

The risk of extinction - the mutational meltdown or the overpopulation.

Science.gov (United States)

Malarz, Krzysztof

2007-04-01

The phase diagrams survival-extinction for the Penna model with parameters: (mutations rate)-(birth rate), (mutation rate)-(harmful mutations threshold), (harmful mutation threshold)-(minimal reproduction age) are presented. The extinction phase may be caused by either mutational meltdown or overpopulation. When the Verhulst factor is responsible for removing only newly born babies and does not act on adults the overpopulation is avoided and only genetic factors may lead to species extinction.

Evaluation of results of Phase B of the German Risk Study for Nuclear Power Plants

International Nuclear Information System (INIS)

Fischer, B.; Hahn, L.; Sailer, M.

1989-01-01

The expert opinion summarizes and discusses the intermediate and the final results of Phase B of the German Risk Study. Emphasis is placed on systems analyses and event scenarios, core melt-down scenarios and effects on the containment, failure of the containment and resulting source terms, radioactivity transport and consequences of accidents, multi-system emergencies and external disturbances, accident management and means of mitigation of damage. (DG) [de

Metallographic post-test investigations for the scaled core-meltdown-experiments FOREVER-1 and -2

International Nuclear Information System (INIS)

Mueller, G.; Boehmert, J.

2000-08-01

FOREVER (Failure Of Reactor Vessel Rentention) experiments have been carried out in order to simulate the behaviour the lower head of a reactor pressure vessel under the conditions of a depressurized core melt down scenario. In particular the creep behaviour and the vessel failure mode have been investigated. Metallographic post test investigations have complemented the experimental programme. Samples of different height positions of the vessel of the FOREVER-C1 and -C2 experiments were metallographically examined and characteristic microstructural appearances were identified. Additionally samples with ineffected microstructure were annealed at different temperatures and cooled by different rates and afterwards investigated. In this way the microstructural effects of the temperature regime, the thermomechanical loads and the environmental attack could be characterized. Remarkable effects were characteristic for the FOREVER-C2 experiment where the highest-loaded region below the welding joint reached temperatures of approx. 1100 C and a strong creep damage occurred. In the FOREVER-C1 experiment creep damage could not be observed and the maximum temperature did not exceed 900 C. Environmental attack generated decarburization and oxidation but the effect was restricted to a narrow surface layer. There was almost no chemical interaction between the oxidic melt and the vessel material. (orig.)

Why Competent Persons Have Meltdowns Working with Troubled Students: A Personal Essay

Science.gov (United States)

Long, Nicholas J.

2010-01-01

How do otherwise competent helpers "lose it" in work with certain troubled children and youth? Drawing on extensive research and practice expertise, this article identifies four causes of these predictable professional "meltdowns"--(1) Caught in the Conflict Cycle; (2) Violation of cherished values and beliefs; (3) Tap-in issues; and (4) Carry-in…

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

International Nuclear Information System (INIS)

Blomquist, C.A.; Ariman, T.; Pierce, R.D.; Pedersen, D.R.

1977-01-01

A description of a meltdown cup to be used in the SLSF in-reactor experiments is presented. Thermal analyses have shown that the cup is capable of containing and cooling the postulated quantities of molten fuel and steel. The basic loadings for stress analyses were defined and failure modes were determined. It was shown that both the maximum bending stress and maximum tangential stress in the Inconel vessel are below the material yield stress. Additionally, the axial stress in the Inconel vessel was found to be negligible. The shear stress in the wire-formed retaining ring is much below the maximum shear stress. Therefore, the meltdown cup is capable of performing its required function

Event course analysis of core disruptive accidents; Ereignisablaufanalyse kernzerstoerender Unfaelle

Energy Technology Data Exchange (ETDEWEB)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-08-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

Investigation of activity release during light water reactor core meltdown

International Nuclear Information System (INIS)

Albrecht, H.; Matschoss, V.; Wild, H.

1978-01-01

A test facility was developed for the determination of activity release and of aerosol characteristics under realistic light water reactor core melting conditions. It is composed of a high-frequency induction furnace, a ThO 2 crucible system, and a collection apparatus consisting of membrane and particulate filters. Thirty-gram samples of a representative core material mixture (corium) were melted under air, argon, or steam at 0.8 to 2.2 bar. In air at 2700 0 C, for example, the relative release was 0.4 to 0.7% for iron, chromium, and cobalt and 4 to 11% for tin, antimony, and manganese. Higher release values of 20 to 40% at lower temperatures (2150 0 C, air) were found for selenium, cadmium, tellurium, and cesium. The size distribution of the aerosol particles was trimodal with maxima at diameters of 0.17, 0.30, and 0.73 μm. The result of a qualitative x-ray microanalysis was that the main elements of the melt were contained in each aerosol particle. Further investigations will include larger melt masses and the additional influence of concrete on the release and aerosol behavior

PWR severe accident mitigation measures, the french point of view

International Nuclear Information System (INIS)

Duco, J.; L'Homme, A.; Queniart, D.

1990-01-01

French studies have early considered the fact that, despite all the precautions taken, the possibility of severe accidents cannot be absolutely excluded; these accidents include core meltdown and a more or less significant loss, at an early or later stage, of the confinement of the radioactive substances in the containment. For a given scenario, one can almost always imagine a more severe scenario by postulating additional failures, but it is obvious that, as the severity of the imagined scenario increases, the probability of its occurrence tends towards zero. However, it does not appear reasonable to attempt to set a probability threshold below which the scenarios should be excluded. First of all, the higher the improbability of the scenarios, the greater the uncertainty in the calculation of their probability, with the result that the calculation is not very meaningful. Secondly, and more importantly, this approach ignores the essential problem of accident situation management. From the outset, French studies have been focused on controlling the development of these situations and mitigating their consequences by means of a series of appropriate actions involving, on the one hand, optimum use of the resources available in the installation during the course of the accident and, on the other hand, the taking of protective measures for the population. To attempt to prevent an initial event to degenerate into a severe accident leading to core meltdown if the proper actions are not taken, Electricite de France has proposed a new operating procedure based on the characterization of every possible cooling state of the core

A condensed review of the core catcher in the LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do hee

2001-03-01

The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized

Core catcher concepts future PWR-Plants

International Nuclear Information System (INIS)

Alsmeyer, H.; Werle, H.

1994-01-01

Light water reactors of the next generation should have still greater passive safety, even in the most serious accidents. This includes the long term safe inclusion of the core inventory in the case of core meltdown accidents. The three concepts for cooling the liquefied core outside the reactor pressure vessel examined by KfK should remove the post-shutdown heat by direct contact of the melt with water. The geometric distribution of the melt increases its surface area, so that favourable conditions for heat removal from the poorly thermally-conducting melt are created and complete quick solidification occurs. The experiments examine both the relocation and distribution mechanisms of the melt and the reactions occurring when water enters. As strong interaction is possible on direct contact of the melt with water, an important aim is experimental determination and limitation of any resulting mechanical stresses. (orig./HP) [de

A review of the core catcher design in LMR

International Nuclear Information System (INIS)

Lee, Yong Bum; Hahn, Do Hee

2001-08-01

The overwhelming emphasis in reactor safety is on the prevention of core meltdown. Moreover, although there have been several accidents that have resulted in some fuel melting, to date there have been no accidents severe enough to cause the syndrome of core collapse, reactor vessel melt-through, containment penetration, and dispersal into the ground. Nevertheless, a number of proposals have been made for the design of core catcher systems to control or stop the motion of the molten core mass should such an accident take place. Core catchers may differ in both their location within the reactor system and in the mechanism that is used to cool and control the motion of the core debris. In this report the classification, configuration and main features of the core catcher are described. And also, The core catcher design technologies and processes are presented. Finally the core catcher provisions in constructed and planned LMRs (Liquid Metal Reactors) are summarized and the preliminary assessment on the core catcher installation in KALIMER is presented

Release of fission and activation products during LWR core meltdown

International Nuclear Information System (INIS)

Albrecht, H.; Matschoss, V.; Wild, H.

1978-01-01

Experiments are described by which activity release fractions and aerosol characteristics were investigated for various core melting conditions. Samples of corium and fissium were heated by induction to temperatures of 2800 0 C under air, argon and steam. Release values are presented for Cr, Mn, Fe, Co, Se, Zr, Mo, Cd, Sn, Sb, Te, J, Cs and U. The deposition behaviour of the released products was found to depend strongly on the volatility and on the gas flow rate. Preliminary results of additional measurements indicate that the size distribution of the aerosol particles is trimodal. (author)

Melt propagation in dry core debris beds

International Nuclear Information System (INIS)

Dosanjh, S.S.

1989-01-01

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

A theory of modern cultural shifts and meltdowns.

Science.gov (United States)

Hochberg, Michael E

2004-08-07

Many cultural attributes such as adornment, language slang, mannerisms and rituals are thought to have little or no influence on individual survival and reproduction, functioning rather as markers of cultural identity that promote group cohesion. Here, I show that if cultural markers are under weak selection and subject to loss or substitution, then the breakdown of cultural cohesiveness may proceed without stabilizing reactions until many or most of a culture's identifiers are forever lost. This may culminate in a 'cultural meltdown', whereby the culture is caught in a vortex of ever-decreasing membership and insufficient selection against the accumulation of unfamiliar markers. In progressively altering the topology of communication from diffusion to broadcasting, globalization may be both accelerating the erosion of cultural identities and amplifying dominance behaviours above their normal adaptive levels.

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

International Nuclear Information System (INIS)

Blomquist, C.A.; Pierce, R.D.; Pedersen, D.R.; Ariman, T.

1977-01-01

The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. Thermal analyses were performed with the Argonne-modified version fo the general heat transfer code THTB, based on the instantaneous addition of 3200 0 K molten fuel with a decay heat of 9 W/gm and 1920 0 K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The stress analysis showed that the Inconel vessel would not fail from the pressure loading, it was also shown that brittle fracture of the tungsten liner from thermal gradients is unlikely. Therefore, the melt-down cup meets the structural design requirements. (Auth.)

Core to surge-line energy transport in a severe accident scenario

International Nuclear Information System (INIS)

Marzo, M. di; Almenas, K.; Gopalnarayanan, S.

1994-01-01

The analysis of loss of coolant accidents in a nuclear power plant, which progress to the stage where the core is uncovered, poses important safety related questions. One of these concerns the rate of energy transport to metal components of the primary system. An experimental program has been conducted at the Univ. of Maryland test facility which quantifies the rate of energy transfer from an uncovered core in a B ampersand W (once-through type steam generators) plant. SF 6 is used to simulate the natural circulation driving force of the high pressure steam expected at prototypical conditions. A time-dependent scaling methodology is developed to transpose experimental data to prototypical conditions. To achieve this transformation, a nominal fluid temperature increase rate of 1.0 degrees C/s is inferred from available TMI-2 event data. To bracket the range of potential prototypical transient scenarios, temperature ramps of 0.8 degrees C/s and 1.2 degrees C/s are also considered. Repeated tests, covering a range of test facility conditions, lead to estimated failure times at the surge line nozzle of 1.5 to 2 hours after initiation of the natural circulation phase of the transient

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

International Nuclear Information System (INIS)

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

1986-01-01

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

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.)

KATS experiments to simulate corium spreading in the EPR core catcher concept

International Nuclear Information System (INIS)

Eppinger, B.; Fieg, G.; Schuetz, W.; Stegmaier, U.

2001-01-01

In future Light Water Reactors special devices (core catchers) might be required to prevent containment failure by basement erosion after reactor pressure vessel melt-through during a core meltdown accident. Quick freezing of the molten core masses is desirable to reduce release of radioactivity. Several concepts of core catcher de-vices have been proposed based on the spreading of corium melt onto flat surfaces with subsequent cooling by flooding with water. Therefore a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out 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. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were chemically inert ceramic layers, dry concrete and concrete with a shallow water layer on top. (authors)

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

Preliminary design of a borax internal core-catcher for a gas cooled fast reactor

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Schumacher, G.

1976-09-01

Preliminary thermal calculations show that a core-catcher appears to be feasible, which is able to cope with the complete meltdown of the core and blankets of a 1,000 MWe GCFR. This core-catcher is based on borax (Na 2 B 4 O 7 ) as dissolving material of the oxide fuel and of the fission products occuring in oxide form. The borax is contained in steel boxes forming a 2.1 meter thick slab on the base of the reactor cavity inside the prestressed concrete reactor vessel, just underneath the reactor core. The fission products are dispersed in the pool formed by the liquid borax. The heat power density in the pool is conveniently reduced and the resulting heat fluxes at the borders of the pool can be safely carried away through the PCRV liner and its water cooling system. (orig.) [de

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

Theoretical and experimental methods to determine the properties of molten core components and reaction products. Pt. 2

International Nuclear Information System (INIS)

Nazare, S.; Ondracek, G.; Schulz, B.

1975-10-01

In the course of a loss of coolant accident, a sequence of events would be initiated that ultimately could lead to core melting. The course of these events and the consequences of core meltdown would in part be determined by the properties of the core materials and the products of their interaction. On the basis of available theoretical and experimental results, the report attempts an estimation of properties such as: 1) work of adhesion between UO 2 - and (U,Zr) liquid phase, 2) heat of fusion of some melts, 3) heat capacity of liquid reaction products, 4) viscosity of liquid reaction products, 5) thermal conductivity of liquid reaction products. Experimental work is suggested for those cases, where the estimates need to be improved or verified. (orig.) [de

A assessment of loss-of-heat-sink accident with scram in the LMFBR

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Pratt, W.T.; Sun, Y.H.

1978-01-01

A description of a slow core meltdown in a liquid metal fast breeder reactor is presented for conditions of loss-of-heat-sink following neutronic shutdown. Simple models are developed for the prediction of phase changes and/or relocation of the core materials including fuel, clad, ducts, control rod absorber material (B 4 C), and plenum gases. The sequence of events is accounted for and the accident progression is described up to the point of recriticality. The neutronic behavior of the disrupted core is analyzed in R-Z geometry with a static transport theory code. For most scenarios assessed, the reactor is expected to become recritical although large ramp rates are not anticipated. (author)

Assessment of the loss-of-heat-sink accident with scram in the LMFBR

International Nuclear Information System (INIS)

Bari, R.A.; Ludewig, H.; Pratt, W.T.; Sun, Y.H.

1978-01-01

A description of a slow core meltdown in a liquid metal fast breeder reactor is presented for the conditions of loss-of-heat-sink following neutronic shutdown. Simple models are developed for the prediction of phase changes and/or relocation of the core materials including fuel, clad, ducts, control rod absorber material (B 4 C), and plenum gases. The sequence of events is accounted for and the accident progression is described up to the point of recriticality. The neutronic behavior of the disrupted core is analyzed in R-Z geometry with a static transport theory code. For most scenarios assessed, the reactor is expected to become recritical although large ramp rates are not anticipated

Tchernobyl accident

International Nuclear Information System (INIS)

1986-06-01

First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given [fr

Tests on the release of fission and activation products during core meltdown

Energy Technology Data Exchange (ETDEWEB)

Albrecht, H; Krause, W; Wild, H

1976-01-01

The first available results are related to tests in which the release of the main components of the core melt, namely the steel, zircaloy and uranium components, was determined using ThO/sub 2/ crucibles. The release products are dispersed onto the pipe walls of the transport system and the measuring filters which were installed at about 1 m distance from the melt crucibles. Of these, only the precipitates on the filters have been analyzed so far. In the tests under air, the release was clearly dependent on the maximum temperature reached. The release values for Mo and Mn were the highest with 5-10%; uranium with 0.1% on the other hand, was the lowest. In a steam atmosphere over the melt, the analysis of the filter precipitates for all elements gave considerably lower values than with the tests in air.

Steam explosion - physical foundations and relation to nuclear reactor safety

International Nuclear Information System (INIS)

Schumann, U.

1982-08-01

'Steam explosion' means the sudden evaporation of a fluid by heat exchange with a hotter material. Other terms are 'vapour explosion', 'thermal explosion', and 'energetic fuel-coolant interaction (FCI)'. In such an event a large fraction of the thermal energy initially stored in the hot material may possibly be converted into mechanical work. For pressurized water reactors one discusses (e.g. in risk analysis studies) a core melt-down accident during which molten fuel comes into contact with water. In the analysis of the consequences one has to investigate steam explosions. In this report an overview over the state of the knowledge is given. The overview is based on an extensive literature review. The objective of the report is to provide the basic knowledge which is required for understanding of the most important theories on the process of steam explosions. Following topics are treated: overview on steam explosion incidents, work potential, spontaneous nucleation, concept of detonation, results of some typical experiments, hydrodynamic fragmentation of drops, bubbles and jets, coarse mixtures, film-boiling, scenario of a core melt-down accident with possible steam-explosion in a pressurized water reactor. (orig.) [de

The consequences from liquid pathways after a reactor meltdown accident

International Nuclear Information System (INIS)

Niemczyk, S.J.; Adams, K.G.; Murfin, W.B.; Ritchie, L.T.; Eppel, E.W.; Johnson, J.D.

1981-06-01

The potential radiological impact of a core-melt accident on the human population has been investigated. In particular, the radiation dose received from radioactivity which could reach the population via liquid pathways has been considered. Radioactivity could be released directly to the hydrosphere after a core-melt accident as a result of melt-through of the containment basemat followed by any of three processes: (1) leaching of the melt debris; 2 escape of sumpwater through the hole formed by the melt (or from passage out of the containment by an alternate route); and 3) depressurization of the containment atmosphere through the melt hole. The three types of releases would differ primarily in their rates, their magnitudes and their radioactive compositions. Both the containment atmosphere and the sumpwater releases would occur relatively rapidly. However, most of the radionuclides present in these two releases in substantial quantities would be expected to be rather short-lived. Therefore, such releases could have a significant impact at a specific site only if the travel times of the important radionuclides to the human population were small. In contrast, leaching of radionuclides from the melt debris would be expected to occur relatively slowly. Most of the long-lived isotopes would be expected to be found primarily in the melt debris. Consequently, even though this release occurred relatively slowly, the impact could still be significant. In contrast to the situation for releases to the atmosphere, accidents corresponding to the most probable RSS (Reactor Safety Study) meltdown categories would result in the largest releases to the hydrosphere. Furthermore, substantial amounts of radioactivity would generally be expected to be released to the hydrosphere during any meltdown accident involving complete melt-through of the containment basemat. On the basis of subsurface hydrologies alone, sites range from those that essentially preclude any impacts to the human

COPING WITH GLOBAL MELTDOWN: INDIA’S EXTERNAL SECTOR

Directory of Open Access Journals (Sweden)

Nandita Sethi

2011-07-01

Full Text Available n 2008 the world economy landed itself in the midst of the most severe financial recession which originated in the developed countries and spread to the developing countries which were hit hard through capital reversals, rising borrowing costs, collapsing world trade and commodity prices, and subsiding remittances. India’s engagement with the global economy became deeper from the 1990s and this made it inevitable to be affected by the global meltdown. This paper attempts to analyze the impact of the global recession on the external sector of the Indian economy. A pre and post recession (2008 analysis is undertaken to see the impact on trade, capital flows, exchange rates and foreign exchange reserves and external debt. The analysis reveals that while India was adversely impacted due to the slowdown in trade, capital flows and outsourcing; the impact has not been as much as in many other Asian economies. The capital flows were soon back, as India seemed like a save haven for funds with reasonable growth and interest rates. The Indian banking and regulatory system has been credited for following prudential norms and India’s huge domestic market and fiscal stimulus have got the economy almost back on track with growth rates looking up again.

Cooling methods of station blackout scenario for LWR plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. From the RELAP5 code analysis, it was shown that the primary system cooling was practicable by using the turbine-driven auxiliary feed water system. (author)

Cooling methods of station blackout scenario for LWR plants

International Nuclear Information System (INIS)

2012-01-01

The objective of this study is to analyze the cooling method of station blackout scenario for both the BWR and PWR plants by RELAP5 code and to check the validity of the cooling method proposed by the utilities. In the BWR plant cooling scenario, the Reactor Core Isolation Cooling System (RCIC), which is operated with high pressure steam from the reactor, injects cooling water into the reactor to keep the core water level. The steam generated in the core is released into the suppression pool at containment vessel to condense. To restrict the containment vessel pressure rising, the ventilation from the wet-well is operated. The scenario is analyzed by RELAP5 and CONTEMPT-LT code. In the PWR plant scenario, the primary pressure is decreased by the turbine-driven auxiliary feed water system operated with secondary side steam of the steam generators (SGs). And the core cooling is kept by the natural circulation flow at the primary loop. The analytical method of un-uniform flow behavior among the SG U-tubes, which affects the natural circulation flow rate, is developed. (author)

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

International Nuclear Information System (INIS)

Frid, W.E.

1986-01-01

A model has been developed to calculate the expansion and fragmentation of a corium jet, due to the evolution of dissolved gas, during a postulated core meltdown accident. Parametric calculations have been performed for a PWR high pressure accident scenario. Jet breakup occurs within a few jet diameters from the RPV. The diameter of the fragmented jet at the level of the reactor cavity floor is predicted to be 40-130 times the discharge diameter. Particles generated by fragmentation of corium melt are predicted to be in the 30-150 μm size range

Core failure accident pathways and ways to control it

International Nuclear Information System (INIS)

Mayinger, F.

1982-01-01

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

Ecological and toxicological aspects of the partial meltdown of the Chernobyl nuclear power plant reactor

Science.gov (United States)

Eisler, Ronald; Hoffman, David J.; Rattner, Barnett A.; Burton, G. Allen; Cairns, John

1995-01-01

the partial meltdown of the 1000-MW reactor at Chernobyl, Ukraine, on April 26, 1986, released large amounts of radiocesium and other radionuclides into the environment, causing widespread radioactive contamination of Europe and the former Soviet Union.1-7 At least 3,000,000 trillion becquerels (TBq) were released from the fuel during the accident (Table 24.1), dwarfing, by orders of magnitude, radiation released from other highly publicized reactor accidents at Windscale (U.K.) and three-Mile Island (U.S.)3,8 The Chernobyl accident happened while a test was being conducted during a normal scheduled shutdown and is attributed mainly to human error.3

Study of an hypothetical reactor meltdown accident for a 50 MW sub(th) fast reactor

International Nuclear Information System (INIS)

Azevedo, E.M. de.

1983-01-01

A melhodology for determining the energy released in hypothetical reactor meltdown accidents is presented. A numerical code was developed based upon the Nicholson method for a uniform and homogeneous reactor with spherical geometry. A comparative study with other know programs in the literature which use better approximations for small energy released, shows that the methodology used were compatible with those under comparison. Besides the influence of some parameters on the energy released, such as the initial power level and the prompt neutron lifetime was studied under this metodology and its result exhibitted. The Doppler effect was also analyzed and its influence on the energy released has been emphasized. (Author) [pt

Experimental observations of boric acid precipitation scenarios

Energy Technology Data Exchange (ETDEWEB)

Vaghetto, R., E-mail: r.vaghetto@tamu.edu; Childs, M., E-mail: masonchilds@tamu.edu; Jones, P., E-mail: pgjones87@tamu.edu; Lee, S., E-mail: sayalee@tamu.edu; Kee, E., E-mail: erniekee@gmail.com; Hassan, Y.A., E-mail: y-hassan@tamu.edu

2017-02-15

During a Loss of Coolant Accident (LOCA) in Light Water Reactors (LWR), borated water is injected into the core through the safety injection system. The continuous vaporization of the water from the core may increase the concentration of boric acid in the core that, under certain conditions may reach the solubility limit and precipitate. This includes scenarios where the liquid water supply to the core is affected by possible blockages due to debris accumulation. Questions have been raised on the effects of the precipitate in the core on the flow behavior, including the possibility of additional blockages produced by precipitate accumulation. A simple experimental facility was constructed to perform experimental observations of the behavior of borated water under the combined effects of the boiling and the boric acid precipitation (BAP). The facility consists of a transparent polycarbonate vertical pipe where forty-five heated rods have been installed to supply the power to the water to reach the saturation temperature, and maintain a desired boil-off rate. The layout and geometry of the experimental apparatus were conceived to emulate a simplified core of a Pressurized Water Reactor (PWR). Experimental observations have been conducted under two different conditions. Preliminary tests were conducted to observe the behavior of the water and the boric acid precipitate during a boil-off scenario without borated water addition (decreasing water level). During the main test runs, borated water was constantly injected from the top of the test section to maintain a constant mixture level in the test section. Both tests assumed no flow from the bottom of the test section which may be the case of PWR LOCA scenarios in presence of debris-generated core blockage. The observations performed with a set of cameras installed around the test section showed interesting effects of the vapor bubbles on the boric acid precipitate migration and accumulation in the test section. The

Annual meeting on nuclear technology 1982. Technical meeting: Possibilities and effects of serious reactor accidents

International Nuclear Information System (INIS)

1982-01-01

A critical examination of the forecast of a design basis accident, the view of the Sandia National Laboratory on the probability of a steam explosion after a core meltdown accident is comparison with WASH-1400, the possibilities of interactions with the containment structure and fission product release, as well as the influences for the assessment of risk in Germany taken from the analysis of core meltdown accidents are dealt with in these papers. (DG) [de

Experimental investigations of the meltdown phase of UO2-Zircaloy fuel rods under conditions of failure of emergency cooling

International Nuclear Information System (INIS)

Hagen, S.; Mack, A.; Malauschek, H.; Wallenfels, K.

1975-01-01

In the monoxidizing helium atmosphere at 1,850 0 C Zircaloy and UO 2 interact violently. The result is a combined meltdown of pellets and can. This phenomenon appears independent of the velocity of temperature rise. In air the oxid skin splits open at 1,890 0 C and the earlier molten material of the interior begins to flow out. When heating up to more than 2,200 0 C the oxid skin remains solid nevertheless. (orig.) [de

The consequences from liquid pathways after a reactor meltdown accident

Energy Technology Data Exchange (ETDEWEB)

Niemczyk, S J; Adams, K G; Murfin, W B; Ritchie, L T; Eppel, E W; Johnson, J D

1981-06-01

The potential radiological impact of a core-melt accident on the human population has been investigated. In particular, the radiation dose received from radioactivity which could reach the population via liquid pathways has been considered. Radioactivity could be released directly to the hydrosphere after a core-melt accident as a result of melt-through of the containment basemat followed by any of three processes: (1) leaching of the melt debris; 2 escape of sumpwater through the hole formed by the melt (or from passage out of the containment by an alternate route); and 3) depressurization of the containment atmosphere through the melt hole. The three types of releases would differ primarily in their rates, their magnitudes and their radioactive compositions. Both the containment atmosphere and the sumpwater releases would occur relatively rapidly. However, most of the radionuclides present in these two releases in substantial quantities would be expected to be rather short-lived. Therefore, such releases could have a significant impact at a specific site only if the travel times of the important radionuclides to the human population were small. In contrast, leaching of radionuclides from the melt debris would be expected to occur relatively slowly. Most of the long-lived isotopes would be expected to be found primarily in the melt debris. Consequently, even though this release occurred relatively slowly, the impact could still be significant. In contrast to the situation for releases to the atmosphere, accidents corresponding to the most probable RSS (Reactor Safety Study) meltdown categories would result in the largest releases to the hydrosphere. Furthermore, substantial amounts of radioactivity would generally be expected to be released to the hydrosphere during any meltdown accident involving complete melt-through of the containment basemat. On the basis of subsurface hydrologies alone, sites range from those that essentially preclude any impacts to the human

Research work for improving LWR safety

International Nuclear Information System (INIS)

Bork, G.

1993-09-01

The work performed in 1992 for the PSF project centers on various phenomena of severe fuel damage and on selected aspects of a core meltdown accident, relating to aerosol behaviour and filter engineering, and to methods of assessing and minimizing the radiological impacts of a reactor accident. The 1992 task programme of the project included research into extreme load conditions affecting the containment in a core meltdown accident: first results are given of the experiments performed. (orig./HP) [de

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

Science.gov (United States)

Kavun, Yusuf; Karasulu, Muzaffer

2015-07-01

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

Effect of the dietary supplement Meltdown on catecholamine secretion, markers of lipolysis, and metabolic rate in men and women: a randomized, placebo controlled, cross-over study

Directory of Open Access Journals (Sweden)

Fisher-Wellman Kelsey H

2009-08-01

Full Text Available Abstract Background We have recently reported that the dietary supplement Meltdown® increases plasma norepinephrine (NE, epinephrine (EPI, glycerol, free fatty acids (FFA, and metabolic rate in men. However, in that investigation measurements ceased at 90 minutes post ingestion, with values for blood borne variables peaking at this time. It was the purpose of the present investigation to extend the time course of measurement to 6 hours, and to include women within the design to determine if sex differences to treatment exist. Methods Ten men (24 ± 4 yrs and 10 women (22 ± 2 yrs ingested Meltdown® or a placebo, using a randomized, cross-over design with one week separating conditions. Blood samples were collected immediately before supplementation and at one hour intervals through 6 hours post ingestion. A standard meal was provided after the hour 3 collection. Samples were assayed for EPI, NE, glycerol, and FFA. Five minute breath samples were collected at each time for measurement of metabolic rate and substrate utilization. Area under the curve (AUC was calculated. Heart rate and blood pressure were recorded at all times. Data were also analyzed using a 2 (sex × 2 (condition × 7 (time repeated measures analysis of variance, with Tukey post hoc testing. Results No sex × condition interactions were noted for AUC for any variable (p > 0.05. Hence, AUC data are collapsed across men and women. AUC was greater for Meltdown® compared to placebo for EPI (367 ± 58 pg·mL-1·6 hr-1 vs. 183 ± 27 pg·mL-1·6 hr-1; p = 0.01, NE (2345 ± 205 pg·mL-1·6 hr-1 vs. 1659 ± 184 pg·mL-1·6 hr-1; p = 0.02, glycerol (79 ± 8 μg·mL-1·6 hr-1 vs. 59 ± 6 μg·mL-1·6 hr-1; p = 0.03, FFA (2.46 ± 0.64 mmol·L-1·6 hr-1 vs. 1.57 ± 0.42 mmol·L-1·6 hr-1; p = 0.05, and kilocalorie expenditure (439 ± 26 kcal·6 hrs-1 vs. 380 ± 14 kcal·6 hrs-1; p = 0.02. No effect was noted for substrate utilization (p = 0.39. Both systolic and diastolic blood pressure

How to arrest a core meltdown accident (doing nothing); Como detener un accidente con fusion de nucleo (sin hacer nada)

Energy Technology Data Exchange (ETDEWEB)

Baron, Jorge H [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

2000-07-01

In the eventual situation of a severe accident in a nuclear reactor, the molten core is able to relocate inside the pressure vessel. This may lead to the vessel failure, due to the thermal attack of the molten core (at approximation of 3000K) on the vessel steel wall. The vessel failure implies the failure of a very important barrier that contains the radioactive materials generated during the reactor operation, with a significant risk of producing high radiation doses both on operators and on the public. It is expected, for the new generation of nuclear reactors, that these will be required to withstand (by design) a core melt down accident, without the need for an immediate evacuation of the surrounding population. In this line, the use of a totally passive system is postulated, which fulfills the objective of containing the molten core inside the pressure vessel, at low temperature (approximation 1200K) precluding its failure. The conceptual design of a passive in-vessel core catcher is presented in this paper, built up of zinc, and designed for the CAREM-25 nuclear power plant. (author)

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

Directory of Open Access Journals (Sweden)

Kavun Yusuf

2015-01-01

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

Review of the TMI-2 accident evaluation and vessel investigation projects

Energy Technology Data Exchange (ETDEWEB)

Ladekarl Thomsen, Knud

1998-03-01

The results of the TMI-2 Accident Evaluation Programme and the Vessel Investigation Project have been reviewed as part of a literature study on core meltdown and in-vessel coolability. The emphasis is placed on the late phase melt progression, which is of special relevance to the NKS-sponsored RAK-2.1 project on Severe Accident Phenomenology. The body of the report comprises three main sections, The TMI-2 Accident Scenario, Core Region and Relocation Path Investigations, and Lower Head Investigations. In the final discussion, the lower head gap formation mechanism is explained in terms of thermal contraction and fracturing of the debris crust. This model seems more plausible than the MAAP model based on creep expansion of the lower head. (au) 1 tab., 33 ills., 31 refs.

Prediction of Transient Scenarios Using AI After Severe Accident Occurrence

International Nuclear Information System (INIS)

Yoo, Kwae Hwan; Back, Ju Hyun; Na, Man Gyun

2017-01-01

We predicted the core uncovery time, the time that core exist temperature (CET) exceeds 1200 .deg. F, reactor vessel (RV) failure time and containment failure time by using the cascaded support vector regression (SVR) model. The proposed algorithms were trained and verified using the simulation data of MAAP code for the optimized power rector (OPR1000). In this study, we predicted transient scenarios by CSVR. The MAAP code was used to describe the accident situation and the 13 measured signal data was acquired and used. The CSVR model was developed to find out the transient scenarios by using short timeintegrated signals after reactor trip. The results show that the CSVR models can predict the transient scenarios accurately.

Modeling of BWR core meltdown accidents - for application in the MELRPI. MOD2 computer code

Energy Technology Data Exchange (ETDEWEB)

Koh, B R; Kim, S H; Taleyarkhan, R P; Podowski, M Z; Lahey, Jr, R T

1985-04-01

This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing.

Re-assessment of road accident data-analysis policy : applying theory from involuntary, high-consequence, low-probability events like nuclear power plant meltdowns to voluntary, low-consequence, high-probability events like traffic accidents

Science.gov (United States)

2002-02-01

This report examines the literature on involuntary, high-consequence, low-probability (IHL) events like nuclear power plant meltdowns to determine what can be applied to the problem of voluntary, low-consequence high-probability (VLH) events like tra...

Dose estimation for nuclear power plant 4 accident in Taiwan at Fukushima nuclear meltdown emission level.

Science.gov (United States)

Tang, Mei-Ling; Tsuang, Ben-Jei; Kuo, Pei-Hsuan

2016-05-01

An advanced Gaussian trajectory dispersion model is used to evaluate the evacuation zone due to a nuclear meltdown at the Nuclear Power Plant 4 (NPP4) in Taiwan, with the same emission level as that occurred at Fukushima nuclear meltdown (FNM) in 2011. Our study demonstrates that a FNM emission level would pollute 9% of the island's land area with annual effective dose ≥50 mSv using the meteorological data on 11 March 2011 in Taiwan. This high dose area is also called permanent evacuation zone (denoted as PEZ). The PEZ as well as the emergency-planning zone (EPZ) are found to be sensitive to meteorological conditions on the event. In a sunny day under the dominated NE wind conditions, the EPZ can be as far as 100 km with the first 7-day dose ≥20 mSv. Three hundred sixty-five daily events using the meteorological data from 11 March 2011 to 9 March 2012 are evaluated. It is found that the mean land area of Taiwan in becoming the PEZ is 11%. Especially, the probabilities of the northern counties/cities (Keelung, New Taipei, Taipei, Taoyuan, Hsinchu City, Hsinchu County and Ilan County) to be PEZs are high, ranging from 15% in Ilan County to 51% in Keelung City. Note that the total population of the above cities/counties is as high as 10 million people. Moreover, the western valleys of the Central Mountain Range are also found to be probable being PEZs, where all of the reservoirs in western Taiwan are located. For example, the probability can be as high as 3% in the far southern-most tip of Taiwan Island in Pingtung County. This shows that the entire populations in western Taiwan can be at risk due to the shortage of clean water sources under an event at FNM emission level, especially during the NE monsoon period. Copyright © 2016 The Authors. Published by Elsevier Ltd.. All rights reserved.

Low void effect (CFV) core concept flexibility: from self-breeder to burner core - 15091

International Nuclear Information System (INIS)

Buiron, L.; Dujcikova, L.

2015-01-01

In the frame of the French strategy on sustainable nuclear energy, several scenarios consider fuel cycle transition toward a plutonium multi-recycling strategy in sodium cooled fast reactor (SFR). Basically, most of these scenarios consider the deployment of a 60 GWe SFR fleet in 2 steps to renew the French PWR fleet. As scenarios do investigate long term deployment configurations, some of them require tools for nuclear phase-out studies. Instead of designing new reactors, the adopted strategy does focus on adaptation of existing ones into burner configurations. This is what was done in the frame of the EFR project at the end of the 90's using the CAPRA approach (French acronym for Enhance Plutonium Consumption in Fast Reactor). The EFR burner configuration was obtained by inserting neutronic penalties inside the core (absorber material and/or diluent subassembly). Starting from the preliminary industrial image of a SFR 3600 MWth core based on Low Sodium Void concept (CFV in French), a 'CAPRA-like' approach has been studied. As the CFV self-breeding is ensured by fertile blankets, a first modification consisted in the substitution of the corresponding depleted uranium by 'inert' or absorber material leading to a 'natural burner' core with only small impacts on flux distribution. The next step forward CAPRA configuration was the substitution of 1/3 of the fuel pins by 'dummy' pins (MgO pellets). The small spectrum shift due to MgO material insertion leads to an increase Doppler constant which exceeds the value of the reference case. As the core sodium void worth value is conserved, the CFV CAPRA core 'safety' potential is quite similar to the one of the reference core. Fuel thermo-mechanical requirements are met by both nominal core power and fuel time residence reduction. However, these reduction factors are lower than those obtained for EFR core. The management of the enhanced reactivity swing is discussed

Preliminary considerations on the startup phase for the ASTRID core

International Nuclear Information System (INIS)

Mignot, G.

2015-01-01

This paper presents preliminary considerations on the startup phase for the ASTRID core, as well as an overview of the different steps before reaching the optimised equilibrium core. The start-up phase is assumed to cover the period between loading the dummy core into the reactor (for commissioning tests) and achieving the optimised equilibrium core. Four main stages are considered: a first stage of start-up tests before fuel core loading, a second stage related to zero power and power ramp-up tests, a third stage corresponding to the transition from the first core to the equilibrium contractual core, and the last stage to reach the optimised performance for the equilibrium core. In the two last stages, a sub-assembly surveillance plan based on post-irradiation examinations is taken into account. As this work is in its preliminary stages, the first scenarios shown for the start-up phase must not be considered as the ASTRID reference scenarios. The scenarios strongly depend on the assumptions considered in the analysis, whereas those discussed in this paper aim at outlining the content and the duration of the starting phases for the ASTRID core, which will be useful in subsequently assessing the core sub-assembly fabrication needs. Assumptions for the start-up phase will be updated in accordance with progress on the ASTRID core design development and core qualification programme. (author)

German risk study nuclear power plants, phase B

International Nuclear Information System (INIS)

1990-01-01

The book documents the results of the probabilistic safety analysis performed at the Biblis-B reference plant. It covers 1) result course analyses and reliability analyses of loss of coolant accidents and transients without and with regard to on-site emergency measures; 2) event tree analyses of on-site and off-site events; 3) analyses of the function of the safety tank in the case of core meltdown accidents, and of the release of fission products into the environment in such case. The probability of a core meltdown is about 3-4 to 100,000 per year and plant. (HP) [de

Numerical analysis of the induction melting process of oxide fuel material

International Nuclear Information System (INIS)

Kondala Rao, R.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

2015-01-01

For the experimental simulation of Molten Fuel-Coolant Interaction (MFCI) phenomenon under hypothetical core meltdown accident scenario in a nuclear reactor, it is required to generate the molten pool of core materials. For this purpose, a laboratory scale Cold wall Crucible induction melting system has been developed. To optimize the system for efficient and reliable melting process, it is required to have comprehensive knowledge on the heat and mass transfer processes along with electromagnetic process that occur during the melting of core materials. Hence, a 2D axi-symmetric numerical model has been developed using a multiphysics software to simulate the induction melting process. The phase change phenomenon is taken into account by using enthalpy formulation. The experimental data available in literature for magnetic field and flow field are used for model validation. The model predicted temperatures are also in good agreement with experimentally measured values. The validated model has been used to study the induction melting behavior of UO_2 fuel material. (author)

Toxicology of plutonium-sodium

International Nuclear Information System (INIS)

Hackett, P.L.

1982-01-01

Scenarios for liquid-metal fast breeder reactor (LMFBR) accidents predict the loss of sodium coolant, with subsequent core melt-down and release of mixed sodium-fuel aerosols [Na-(PuU)O 2 ] into the environment. Studies in other laboratories demonstrated that mixed aerosols of Na 2 O-PuO 2 were more readily transported from the lung than PuO 2 aerosols. We therefore devised a continuous aerosol-generating system for animal exposures in which laser-generated fuel aerosols were swept through sodium vapor to form sodium-fuel aerosols. These fuel and sodium-fuel aerosols were compared with regard to their physicochemical properties and their biological behavior following inhalation studies in rats and dogs

CNE (central nuclear en Embalse): probabilistic safety study. Loss-of-coolant accidents. Analysis through events sequence

International Nuclear Information System (INIS)

Layral, S.I.

1987-01-01

The aim of this study was to perform for the Embalse nuclear power plant, a probabilistic evaluation of loss-of-coolant accidents (LOCA) to identify the risks associated with them and to determine their acceptability in accordance with norms. This study includes all ruptures in the primary system that produce the automatic activation of 'emergency core cooling system'. Three starting events were selected for the probabilistic evaluation: 100% rupture of an input collector; 5% rupture of an input collector; 1.2% rupture of an input collector. At this stage the evaluation is focussed on the identification and quantization of the main failure sequences that follow a LOCA and lead to an uncontrolled reactor state or 'core meltdown'. The most important contribution to the core meltdown due to LOCA is the failure of supplies that are required for the emergency core cooling system. (Author)

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

Energy Technology Data Exchange (ETDEWEB)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

2008-07-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

International Nuclear Information System (INIS)

Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

2008-01-01

The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

International Nuclear Information System (INIS)

Kroeger, P.G.

1988-01-01

Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

Research work for improving LWR safety. 1991 annual report

International Nuclear Information System (INIS)

1992-08-01

The work performed in 1991 for the PNS project centers on various phenomena of severe fuel damage and on selected aspects of a core meltdown accident, relating to aerosol behaviour and filter engineering, and to methods of assessing and minimizing the radiological impacts of a reactor accident. One paper included in this annual report summarizes the evaluation of experiments carried out in 1990 at the Wuerenlingen PROTEUS reactor. The 1991 task programme of the project for the first time included research into extreme load conditions affecting the containment in a core meltdown accident; first results are given of the experiments performed. (orig./DG) [de

Core design and performance of small inherently safe LMRs

International Nuclear Information System (INIS)

Orechwa, Y.; Khalil, H.; Turski, R.B.; Fujita, E.K.

1986-01-01

Oxide and metal-fueled core designs at the 900 MWt level and constrained by a requirement for interchangeability are described. The physics parameters of the two cores studied here indicate that metal-fueled cores display attractive economic and safety features and are more flexible than are oxide cores in adapting to currently-changing deployment scenarios

Exorcising the nightmare of reactor meltdowns. First of four articles

International Nuclear Information System (INIS)

Faltermayer, E.

1979-01-01

Facts on the hazards of reactors are investigated in this first article of four on that subject. Some knowledgeable critics warn that a serious nuclear disaster is a real possibility within the next few decades. Other experts, mostly in the nuclear industry, counter that fission power is just another engineering challenge that has been met by a society that copes with many others. Each year, they note, U.S. industry handles enough chlorine to kill the entire human race 100,000 times over. U.S. utilities have accumulated 440 reactor-years of experience with nonbreeeding reactors with no meltdowns. Beginning in 1972, a study (eventually report WASH-1400) was made of hundreds of sequences of things that could go wrong in commercial reactors, including human error, and estimated the amounts of radioactivity that could be released in each sequence under varying weather conditions; the optimistic analyses in that report are well known. In January 1979, the Nuclear Regulatory Commission issued a statement saying that it does not regard as reliable WASH-1400's figures on the overall risk of reactor accidents. In March 1975, some 1,600 electrical cables were burned in a seven-hour fire at Brown's Ferry nuclear plant near Athens, Alabama with no major accident occurring. .The LOFT reactor experiment in Idaho is described. Pipe cracks are the most worrisome problem, but these get discovered and corrected before an accident occurs. In a society that long ago lost its technological innocence, and that lives with far greater hazards both natural and manmade, nuclear power causes more unease that it should

Scoping Analysis on Core Disruptive Accident in PGSFR (2015 Results)

Energy Technology Data Exchange (ETDEWEB)

Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

2016-05-15

In general, the severe accident is classified by three phases. The first phase is the initiation (pre-disassembly) phase that occurs the gradual core meltdown from accident initiation to the point of neutronic shutdown with an intact geometry. The second phase is the transition phase that happens the fuel transition from a solid to a liquid phase. Fuel and cladding can melt to form a molten pool and core can boil, then criticality conditions can recur. The third phase is the disassembly phase. In other words, this phase is Core Disruptive Accident (CDA). Power excursion is followed until the core is disassembled in this phase. In the early considerations of Liquid Metal Fast Breeder Reactor (LMFBR) energetics, the term Hypothetical Core Disruptive Accidents (HCDAs) was in common use. This was not only to connote the extremely low probability of initiation of such accidents, but also the tentative nature of our understanding of their behavior and resulting consequences. A numerical analysis is conducted to estimate the energy release, pressure behavior and core expansion behavior induced by CDA of PGSFR using CDA-ER and CDA-CEME codes. Conservatively, the calculated results of energy release and pressure behavior induced by CDA without Doppler effect in PGSFR when whole cores were melted (100 $/s) were 7.844 GJ and 4.845 GPa, respectively. With Doppler effect, the analyzed maximum energy release and pressure were 6.696 GJ and 3.449 GPa, respectively. The calculated results of the core expansion behavior during 0.015 seconds after the explosion without Doppler effect in PGSFR when whole cores were melted (100 $/s) were as follows: The total energy is calculated to be 1.87 GJ. At 0.01 s, the kinetic energy of the sodium is 1.85 GJ, while the expansion work and internal energy of the bubble are 19.7 MJ and 0.98 J, respectively. With Doppler effect, the total energy is calculated to be 1.33 GJ. At 0.01 s, the kinetic energy of the sodium is 1.31 GJ, while the expansion

Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Andrews, Nathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Faucett, Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Haskin, Troy Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Luxat, Dave [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Geiger, Garrett [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Codella, Brittany [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-10-01

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.

Development on quantitative safety analysis method of accident scenario. The automatic scenario generator development for event sequence construction of accident

International Nuclear Information System (INIS)

Kojima, Shigeo; Onoue, Akira; Kawai, Katsunori

1998-01-01

This study intends to develop a more sophisticated tool that will advance the current event tree method used in all PSA, and to focus on non-catastrophic events, specifically a non-core melt sequence scenario not included in an ordinary PSA. In the non-catastrophic event PSA, it is necessary to consider various end states and failure combinations for the purpose of multiple scenario construction. Therefore it is anticipated that an analysis work should be reduced and automated method and tool is required. A scenario generator that can automatically handle scenario construction logic and generate the enormous size of sequences logically identified by state-of-the-art methodology was developed. To fulfill the scenario generation as a technical tool, a simulation model associated with AI technique and graphical interface, was introduced. The AI simulation model in this study was verified for the feasibility of its capability to evaluate actual systems. In this feasibility study, a spurious SI signal was selected to test the model's applicability. As a result, the basic capability of the scenario generator could be demonstrated and important scenarios were generated. The human interface with a system and its operation, as well as time dependent factors and their quantification in scenario modeling, was added utilizing human scenario generator concept. Then the feasibility of an improved scenario generator was tested for actual use. Automatic scenario generation with a certain level of credibility, was achieved by this study. (author)

TMI-2 core bore acquisition summary report

International Nuclear Information System (INIS)

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

1986-09-01

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

Recriticality analyses for CAPRA cores

International Nuclear Information System (INIS)

Maschek, W.; Thiem, D.

1995-01-01

The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

Recriticality analyses for CAPRA cores

Energy Technology Data Exchange (ETDEWEB)

Maschek, W.; Thiem, D.

1995-08-01

The first scoping calculation performed show that the energetics levels from recriticalities in CAPRA cores are in the same range as in conventional cores. However, considerable uncertainties exist and further analyses are necessary. Additional investigations are performed for the separation scenarios of fuel/steel/inert and matrix material as a large influence of these processes on possible ramp rates and kinetics parameters was detected in the calculations. (orig./HP)

Analysis of recessive sex-linked lethal mutations in genetically different strains of Drosophila melanogaster ms and w irradiated in the five-kilometer zone of the Chernobyl meltdown

International Nuclear Information System (INIS)

Aslanyan, M.M.; Kim, A.I.; Magomedova, M.A.; Fatkulbayanova, N.L.

1994-01-01

The frequency of induced and spontaneous recessive sex-linked lethal mutations (RSLLM) in Drosophila melanogaster strains w and ms was estimated after their chronic irradiation in the five-kilometer zone of the Chernobyl' meltdown. The mutagenic effect of relatively low radiation doses was analyzed. In an experiment conducted in 1990, a significant increase in the RSLLM frequency was recorded, while, in 1991, no significant difference between the experiment and control was found

Post-accident heat removal research: A state of the art review

International Nuclear Information System (INIS)

Mueller, U.; Schulenberg, T.

1983-11-01

For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

Dose estimation for nuclear power plant 4 accident in Taiwan at Fukushima nuclear meltdown emission level

International Nuclear Information System (INIS)

Tang, Mei-Ling; Tsuang, Ben-Jei; Kuo, Pei-Hsuan

2016-01-01

An advanced Gaussian trajectory dispersion model is used to evaluate the evacuation zone due to a nuclear meltdown at the Nuclear Power Plant 4 (NPP4) in Taiwan, with the same emission level as that occurred at Fukushima nuclear meltdown (FNM) in 2011. Our study demonstrates that a FNM emission level would pollute 9% of the island's land area with annual effective dose ≥50 mSv using the meteorological data on 11 March 2011 in Taiwan. This high dose area is also called permanent evacuation zone (denoted as PEZ). The PEZ as well as the emergency-planning zone (EPZ) are found to be sensitive to meteorological conditions on the event. In a sunny day under the dominated NE wind conditions, the EPZ can be as far as 100 km with the first 7-day dose ≥20 mSv. Three hundred sixty-five daily events using the meteorological data from 11 March 2011 to 9 March 2012 are evaluated. It is found that the mean land area of Taiwan in becoming the PEZ is 11%. Especially, the probabilities of the northern counties/cities (Keelung, New Taipei, Taipei, Taoyuan, Hsinchu City, Hsinchu County and Ilan County) to be PEZs are high, ranging from 15% in Ilan County to 51% in Keelung City. Note that the total population of the above cities/counties is as high as 10 million people. Moreover, the western valleys of the Central Mountain Range are also found to be probable being PEZs, where all of the reservoirs in western Taiwan are located. For example, the probability can be as high as 3% in the far southern-most tip of Taiwan Island in Pingtung County. This shows that the entire populations in western Taiwan can be at risk due to the shortage of clean water sources under an event at FNM emission level, especially during the NE monsoon period. - Highlights: • An advanced Gaussian-type trajectory model to evaluate the evacuation zone at Nuclear Power Plant 4 in Taiwan. • Mean land area of Taiwan in becoming the permanent evacuation zone is 11%. • The probabilities of the northern

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

Energy Technology Data Exchange (ETDEWEB)

Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kondo, Toshiki [Graduate School of Engineering, Osaka University (Japan); Ishikawa, Takehiko [Japan Aerospace Exploration Agency (Japan); SOKEN-DAI (Graduate University for Advanced Studies) (Japan); Okada, Junpei T. [Institute for Materials Research, Tohoku University (Japan); Watanabe, Yuki [Advanced Engineering Services Co. Ltd. (Japan); Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

2017-03-15

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 0.77}Cr{sub 0.23}) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr{sub 1-x}Ni{sub x} (x = 0.12 and 0.24) and Zr{sub 77}Cr{sub 23}. • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

International Nuclear Information System (INIS)

Ohishi, Yuji; Kondo, Toshiki; Ishikawa, Takehiko; Okada, Junpei T.; Watanabe, Yuki; Muta, Hiroaki; Kurosaki, Ken; Yamanaka, Shinsuke

2017-01-01

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 0.77 Cr 0.23 ) using the electrostatic levitation technique. - Highlights: • The physical properties of Zr-Ni and Zr-Cr liquid alloys have been measured Zr 1-x Ni x (x = 0.12 and 0.24) and Zr 77 Cr 23 . • The measurement was conducted using the electrostatic levitation technique. • The density, viscosity, and surface tension of each liquid alloy were measured.

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

Science.gov (United States)

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

2018-03-01

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

Four-fluid model of PWR degraded cores

International Nuclear Information System (INIS)

Dearing, J.F.

1985-01-01

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

A synthesis on the HTR scenario studies at CEA - HTR2008-58059

International Nuclear Information System (INIS)

Boucher, L.; Greneche, D.

2008-01-01

The aim of the studies is to assess the impact of the deployment of an HTR park replacing one part of the current PWR reactors. The other part of the current park is replaced by EPRs. In these scenarios, the annual electricity production is constant at 400 TWhe. This value corresponds roughly to the present nuclear electricity production in France. From 2002 to 2007, an important program study on HTR has been carried out by CEA and AREVA NC under the joint CEA - AREVA NC project 'prospective studies on the management of Plutonium and the back end of the cycle'. This program addresses core physic and scenario studies, and also the back end of the fuel cycle : reprocessing of spent fuel and HTR waste management. Some core physic studies have already been presented in the reference [1]. This paper presents the results of the scenario studies using two concepts: either the standard core of the Gas Turbine Modular Helium Reactor concept (GTMHR) with Uranium or Plutonium fuel, or the Multiple Fuel Rows Core (MFRC) dedicated to the actinide burning. The insertion of a new concept (fuel, reactor, process) must be evaluated in the global electronuclear system with an analysis of the impact on the fuel cycle (Enrichment, Fuel Fabrication, Reactor, Processing, Interim Storage, Waste storage). The scenario studies are used to evaluate different solutions to manage nuclear materials (uranium, plutonium) and wastes (minor actinides and fission products), from the present situation in France (closed cycle with storage of used MOX fuels) until the final equilibrium: mixed nuclear park with EPR and HTR. These studies allow to calculate material flows and inventories of these elements in each step of the fuel cycle. The simulation of transient scenarios from the present situation to the future situation is performed with the COSI code. HTR reactors feature a high flexibility with regard to fuel cycle options. Several versions of core have been investigated, with different type of

Formation and fragmentation of protostellar dense cores

International Nuclear Information System (INIS)

Maury, Anaelle

2009-01-01

Stars form in molecular clouds, when they collapse and fragment to produce protostellar dense cores. These dense cores are then likely to contract under their own gravity, and form young protostars, that further evolve while accreting their circumstellar mass, until they reach the main sequence. The main goal of this thesis was to study the formation and fragmentation of protostellar dense cores. To do so, two main studies, described in this manuscript, were carried out. First, we studied the formation of protostellar cores by quantifying the impact of protostellar outflows on clustered star formation. We carried out a study of the protostellar outflows powered by the young stellar objects currently formed in the NGc 2264-C proto-cluster, and we show that protostellar outflows seem to play a crucial role as turbulence progenitors in clustered star forming regions, although they seem unlikely to significantly modify the global infall processes at work on clump scales. Second, we investigated the formation of multiple systems by core fragmentation, by using high - resolution observations that allow to probe the multiplicity of young protostars on small scales. Our results suggest that the multiplicity rate of protostars on small scales increase while they evolve, and thus favor dynamical scenarios for the formation of multiple systems. Moreover, our results favor magnetized scenarios of core collapse to explain the small-scale properties of protostars at the earliest stages. (author) [fr

Thermal conditions and functional requirements for molten fuel containment

International Nuclear Information System (INIS)

Kang, C.S.; Torri, A.

1980-05-01

This paper discusses the configuration and functional requirements for the molten fuel containment system (MFCS) in the GCFR demonstration plant design. Meltdown conditions following a loss of shutdown cooling (LOSC) accident were studied to define the core debris volume for a realistic meltdown case. Materials and thicknesses of the molten fuel container were defined. Stainless steel was chosen as the sacrificial material and magnesium oxide was chosen as the crucible material. Thermal conditions for an expected quasi-steady state were analyzed. Highlights of the functional requirements which directly affect the MFCS design are discussed

Measures for removing hydrogen

International Nuclear Information System (INIS)

Baukal, W.; Koehling, A.; Langer, G.; Poeschel, E.

1984-01-01

Basis for the investigation is a 1300-MW-PWR. The evolution of hydrogen was studied in design-basis and three hypothetical accident scenarios, the loss-of-coolant accident, the failure of emergency cooling system and core meltdown. It was shown that in the case of release rates of 4m 3 H 2 /h, the known post-accident hydrogen removal systems can be used and at medium rates up to 80 m 3 H 2 /h recombines of nuclear and non-nuclear industries are suitable under certain conditions. In the case of larger release rates it appears useful to apply a small recombiner of the type of the post-accident hydrogen removal system combined with an other hydrogen countermeasures. Recommendations are being made for the installation of an accident-proof hydrogen measuring system. (DG) [de

Assessment of Mass Fraction and Melting Temperature for the Application of Limestone Concrete and Siliceous Concrete to Nuclear Reactor Basemat Considering Molten Core–Concrete Interaction

Directory of Open Access Journals (Sweden)

Hojae Lee

2016-04-01

Full Text Available Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core–concrete interaction analysis.

Loss of coolant accident mitigation for liquid metal cooled space reactors

International Nuclear Information System (INIS)

Georgevich, Vladimir; Best, Frederick; Erdman, Carl

1989-01-01

A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

Severe accident analysis to prevent high pressure scenarios in the EPR TM

International Nuclear Information System (INIS)

Azarian, G.; Gandrille, P.; Gasperini, M.; Klein, R.

2010-01-01

The EPR TM has incorporated several design features in order to specifically address major severe accident safety issues. In particular, it was designed with the objective to transfer high pressure core melt scenarios into a low pressure scenario with high reliability so that a high pressure vessel failure can be practically eliminated. It is the key issue in the defense-in-depth approach, for a postulated severe accident with core melting, to prevent any risk of containment failure due to possible Direct Containment Heating or due to reactor vessel rocketing which results from vessel failure at high pressure. Temperature-induced steam generator tube rupture, which could lead to a radiological containment bypass, has also to be prevented. On the basis of the analysis of the main high pressure core melt scenarios which are calculated with the MAAP4.07 code which was developed to support the EPR TM, this paper explores the benefits of primary depressurization by dedicated valves on transient evolutions. It specifically addresses the thermal response of the structures by sensitivity studies involving the timing of valve actuation. It outlines that a grace period of at least one hour is available for a delayed valve actuation without inducing excessive loads and without increasing the risk of a temperature-induced steam generator tube rupture. (authors)

Main results of German risk study, Phase B, in the light of methodological problems

International Nuclear Information System (INIS)

Koeberlein, K.

1991-01-01

The German Risk Study, Phase B, which has been performed for a 1,300 MWe KWU-type pressurized water reactor, is a level 2 probabilistic safety assessment (PSA). In the study, 32 initiating events have been considered. An expected plant damage state frequency of about 3x10 -5 per year has been calculated. In a plant damage state core meltdown can only be prevented by means of accident management procedures. The probability is high (about 98%) that in a plant damage state the primary system will not be depressurized. Core meltdown under high system pressure would endanger the containment integrity immediately. In most plant damage states accident management procedures are possible which bring the plant in a safe state or, at least, reduce the pressure in the primary system before the pressure vessel fails. On a preliminary basis, a success probability of 99% for accident management procedures (secondary or primary side bleed and feed) has been estimated. This results in an expected core melt frequency of about 4x10 -6 per year. The expected frequency of high pressure core meltdown is about 6x10 -7 per year. Loads on, and failure modes of, the containment during a core melt accident as well as release fractions after containment failure have been analyzed. The study did, however, refrain from quantifying the probability of containment failure modes. A main uncertainty in this area is the probability of containment failure due to hydrogen burning. The results of the study point out, that the risk can be reduced significantly if containment failure due to hydrogen burning can be reliably prevented

Severe accident considerations for modern KWU-PWR plants

International Nuclear Information System (INIS)

Eyink, J.

1987-01-01

In assumption of severe accident on modern KWU-PWR plants the author discusses on the: selection of core meltdown sequences, course of the accident, containment behaviour and source terms for fission products release to the environment

Poison and diluent system for nuclear power plants

International Nuclear Information System (INIS)

Parker, W.G.; Ravets, J.M.; Preble, B.S.

1978-01-01

A system to prevent supercriticality in nuclear power plants in the unlikely event of a core destructive accident terminating in the nuclear core meltdown is described. The system dilutes and poisons the molten core to maintain subcriticality, and is useful in mobile nuclear power plants, or in nuclear plants subject to seismic disturbances, where the orientation of the nuclear reactor after the accident is unknown. It is also applicable to alleviate the consequences of loss of coolant flow accidents from any cause. Aside from preventing supercriticality, the system serves the dual purpose of acting as a biological shield and/or structural member that reduces the deleterious effects of accidental core impaction, without compromising power plant weight and size constraints. A borated material, with a melting point greater than the fuel melting point, is inserted in the pressure vessel behind an inner wall. In the unlikely event of a core meltdown, the molten fuel melts through the inner wall and is diluted and poisoned by the borated material. In the event the molten fuel melts through the pressure vessel, additional borated material is provided to continue diluting and poisoning

Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

Science.gov (United States)

Valentin Rodriguez, Francisco Ivan

High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

Based on user interest level of modeling scenarios and browse content

Science.gov (United States)

Zhao, Yang

2017-08-01

User interest modeling is the core of personalized service, taking into account the impact of situational information on user preferences, the user behavior days of financial information. This paper proposes a method of user interest modeling based on scenario information, which is obtained by calculating the similarity of the situation. The user's current scene of the approximate scenario set; on the "user - interest items - scenarios" three-dimensional model using the situation pre-filtering method of dimension reduction processing. View the content of the user interested in the theme, the analysis of the page content to get each topic of interest keywords, based on the level of vector space model user interest. The experimental results show that the user interest model based on the scenario information is within 9% of the user's interest prediction, which is effective.

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

International Nuclear Information System (INIS)

Phung, Viet-Anh; Galushin, Sergey; Raub, Sebastian; Goronovski, Andrei; Villanueva, Walter; Kööp, Kaspar; Grishchenko, Dmitry; Kudinov, Pavel

2016-01-01

Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small ( 100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small variations of the input

Characteristics of debris in the lower head of a BWR in different severe accident scenarios

Energy Technology Data Exchange (ETDEWEB)

Phung, Viet-Anh, E-mail: vaphung@kth.se; Galushin, Sergey, E-mail: galushin@kth.se; Raub, Sebastian, E-mail: raub@kth.se; Goronovski, Andrei, E-mail: andreig@kth.se; Villanueva, Walter, E-mail: walterv@kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

2016-08-15

Highlights: • Station blackout scenario with delayed recovery of safety systems in a Nordic BWR is considered. • Genetic algorithm and random sampling methods are used to explore accident scenario domain. • Main groups of scenarios are identified. • Ranges and distributions of characteristics of debris bed in the lower head are determined. - Abstract: Nordic boiling water reactors (BWRs) adopt ex-vessel debris cooling to terminate severe accident progression. Core melt released from the vessel into a deep pool of water is expected to fragment and form a coolable debris bed. Characteristics of corium melt ejection from the vessel determine conditions for molten fuel–coolant interactions (FCI) and debris bed formation. Non-coolable debris bed or steam explosion can threaten containment integrity. Vessel failure and melt ejection mode are determined by the in-vessel accident progression. Characteristics (such as mass, composition, thermal properties, timing of relocation, and decay heat) of the debris bed formed in the process of core relocation into the vessel lower plenum define conditions for the debris reheating, remelting, melt-vessel structure interactions, vessel failure and melt release. Thus core degradation and relocation are important sources of uncertainty for the success of the ex-vessel accident mitigation strategy. The goal of this work is improve understanding how accident scenario parameters, such as timing of failure and recovery of different safety systems can affect characteristics of the debris in the lower plenum. Station blackout scenario with delayed power recovery in a Nordic BWR is considered using MELCOR code. The recovery timing and capacity of safety systems were varied using genetic algorithm (GA) and random sampling methods to identify two main groups of scenarios: with relatively small (<20 tons) and large (>100 tons) amount of relocated debris. The domains are separated by the transition regions, in which relatively small

Exploratory study of molten core material/concrete interactions, July 1975--March 1977

International Nuclear Information System (INIS)

Powers, D.A.; Dahlgren, D.A.; Muir, J.F.; Murfin, W.D.

1978-02-01

An experimental study of the interaction between high-temperature molten materials and structural concrete is described. The experimental efforts focused on the interaction of melts of reactor core materials weighing 12 to 200 kg at temperatures 1700 to 2800 0 C with calcareous and basaltic concrete representative of that found in existing light-water nuclear reactors. Observations concerning the rate and mode of melt penetration into concrete, the nature and generation rate of gases liberated during the interaction, and heat transfer from the melt to the concrete are described. Concrete erosion is shown to be primarily a melting process with little contribution from mechanical spallation. Water and carbon dioxide thermally released from the concrete are extensively reduced to hydrogen and carbon monoxide. Heat transfer from the melt to the concrete is shown to be dependent on gas generation rate and crucible geometry. Interpretation of results from the interaction experiments is supported by separate studies of the thermal decomposition of concretes, response of bulk concrete to intense heat fluxes (28 to 280 W/cm 2 ), and heat transfer from molten materials to decomposing solids. The experimental results are compared to assumptions made in previous analytic studies of core meltdown accidents in light-water nuclear reactors. A preliminary computer code, INTER, which models and extrapolates results of the experimental program is described. The code allows estimation of the effect of physical parameters on the nature of the melt/concrete interaction

Development of a research simulator for the study of human factors and experiments

International Nuclear Information System (INIS)

Kawano, R.; Shibuya, S.

1999-01-01

A research simulator of nuclear power plant for Human Factors was developed. It simulates the behaviors of the 1100MWe BWR nuclear power plant and has almost same functions ant scope of the simulation as a full-scope training simulator. Physical models installed in the system enable us to execute experiments with multi-malfunction scenario. A severe accident simulation package replaces the running simulation code when the maximum core temperature exceeds 1200 deg C and the core approaches meltdown conditions. The central control panel was simulated by soft panels, indicator and operational switches on the panels by computer graphics, displayed on 22 console boxes containing CRT. The introduction of soft panels and EWSs connected with LAN accomplished flexibility and extendibility. Some experiments by using the simulator were executed and the system has been improved based on the experience from the experiments. It is important to evaluate the effectiveness of any new system by using an actual plant size research simulator before its practical application to keep steady and safe operation of nuclear power plants. (author)

Severe accidents in nuclear reactors

International Nuclear Information System (INIS)

Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

2004-01-01

The likelihood of accidents leading to core meltdown in nuclear reactors is low. The consequences of such an event are but so severe that developing and implementing of adequate measures for preventing or diminishing the consequences of such events are of paramount importance. The analysis of major accidents requires sophisticated computation codes but necessary are also relevant experiments for checking the accuracy of the predictions and capability of these codes. In this paper an overview of the severe accidents worldwide with definitions, computation codes and relating experiments is presented. The experimental research activity of severe accidents was conducted in INR Pitesti since 2003, when the Institute jointed the SARNET Excellence Network. The INR activity within SARNET consists in studying scenarios of severe accidents by means of ASTEC and RELAP/SCDAP codes and conducting bench-scale experiments

Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

International Nuclear Information System (INIS)

Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

2015-01-01

Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Passive decay heat removal by sump cooling after core meltdown

International Nuclear Information System (INIS)

Knebel, J.U.; Mueller, U.

1996-01-01

This article presents the basic physical phenomena and scaling criteria of decay heat removal from a large coolant pool by single-phase and two-phase natural circulation flow. The physical significance of the dimensionless similarity groups derived is evaluated. The above results are applied to the SUCO program that is performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. The sump cooling concept is based on passive safety features within the containment. The work is supported by the German utilities and the Siemens AG. The article gives first measurement results of the 1:20 linearly scaled plane two-dimensional SUCOS-2D test facility. The experimental results of the model geometry are transformed to prototype conditions

Transuranium contamination in BWRs after fuel accidents and its impact on decommissioning exposures and costs

Energy Technology Data Exchange (ETDEWEB)

Lundgren, K.

1996-12-01

The theme of the present study is to quantify the amount of transuranium activity in different parts of the plant after various fuel accidents, and which impact such contamination has on radiation exposure and costs for decommissioning the plant. The consequences of four different accident degrees have been treated: Common fuel failures, e.g. in line with recent experiences from Swedish BWRs; Fuel channel obstruction resulting in partial melting of one fuel assembly; Total loss of electric power resulting in partial meltdown of the core, but with primary circuit intact preventing a massive contamination of the containment; A LOCA followed by a core meltdown and melting and penetration of the reactor pressure vessel. The amount of transuranium activity distributed, the form of this activity and the plant contamination are evaluated for these accidents. The costs and exposures have been split up on cleanup activities after the accident and decommissioning. 75 refs.

Transuranium contamination in BWRs after fuel accidents and its impact on decommissioning exposures and costs

International Nuclear Information System (INIS)

Lundgren, K.

1996-12-01

The theme of the present study is to quantify the amount of transuranium activity in different parts of the plant after various fuel accidents, and which impact such contamination has on radiation exposure and costs for decommissioning the plant. The consequences of four different accident degrees have been treated: Common fuel failures, e.g. in line with recent experiences from Swedish BWRs; Fuel channel obstruction resulting in partial melting of one fuel assembly; Total loss of electric power resulting in partial meltdown of the core, but with primary circuit intact preventing a massive contamination of the containment; A LOCA followed by a core meltdown and melting and penetration of the reactor pressure vessel. The amount of transuranium activity distributed, the form of this activity and the plant contamination are evaluated for these accidents. The costs and exposures have been split up on cleanup activities after the accident and decommissioning. 75 refs

Advanced scenarios for ITER operation

Energy Technology Data Exchange (ETDEWEB)

Sips, A.C.C. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

2004-07-01

In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of {alpha}-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in

Towards a user's guide to scenarios - a report on scenario types and scenario techniques

Energy Technology Data Exchange (ETDEWEB)

Boerjeson, Lena; Hoejer, Mattias; Dreborg, Karl-Henrik; Finnveden, Goeran [Royal Inst. of Technology, Stockholm (Sweden). Environmental Strategies Research - fms; Ekvall, Tomas [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Energy and Environment

2005-11-01

Futures studies consist of a vast variation of studies and approaches. The aim of this paper is to contribute to the understanding of for what purposes scenarios are useful and what methods and procedures are useful for furthering these purposes. We present a scenario typology with an aim to better suit the context in which the scenarios are used. The scenario typology is combined with a new way of looking at scenario techniques, i.e. practical methods and procedures for scenario development. Finally, we look at the usefulness of scenarios in the light of the scenario typology and the scenario techniques. As a start, we distinguish between three main categories of scenario studies. The classification is based on the principal questions we believe a user may want to pose about the future. The resolution is then increased by letting each category contain two different scenario types. These are distinguished by different angles of approach of the questions defining the categories. The first question, What will happen?, is responded to by Predictive scenarios. In fact, the response to a question like this will always be conditional, e.g. of a stable and peaceful world, or by a certain continuous development of some kind. We have utilized this fact when defining the two predictive scenario types, Forecasts and What-if scenarios. The second question, What can happen?, is responded to by Explorative scenarios. The scenarios are thus explorations of what might happen in the future, regardless of beliefs of what is likely to happen or opinions of what is desirable. This category is further divided into external and strategic scenarios. The final question, How can a specific target be reached?, is responded to by Normative scenarios. Such studies are explicitly normative, since they take a target as a starting point. They are often directed towards how the target could be reached. This category is divided into preserving and transforming scenarios. If the user wants to

Applicability of Phebus FP results to severe accident safety evaluations and management measures

International Nuclear Information System (INIS)

Schwarz, M.; Clement, B.; Jones, A.V.

2001-01-01

The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment

List of the reports from reactor safety research of the BMFT and USNRC

International Nuclear Information System (INIS)

1977-02-01

This list presents a survey of reports from the FRG and USA issued on special problems of reactor safety research. The problems include emergency core cooling, the containment during loss of coolant, core meltdown, component safety, external impacts, risk and reliability, quality assurance, fission product transport, radiation exposure, as well as the LWR, HTR and the fast sodium-cooled reactor in general. (orig./HK) [de

Current position on severe accident phenomena

Energy Technology Data Exchange (ETDEWEB)

Henry, Robert E [Fauske and Associates, Inc., Burr Ridge, IL (United States)

2004-07-01

The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors.

Current position on severe accident phenomena

International Nuclear Information System (INIS)

Henry, Robert E.

2004-01-01

The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors

Multi-Scenario Use Case based Demonstration of Buildings Cybersecurity Framework Webtool

Energy Technology Data Exchange (ETDEWEB)

Gourisetti, Sri Nikhil G.; Mylrea, Michael E.; Gervais, Easton L.; Bhadra, Sraddhanjoli

2017-11-27

The purpose of this paper is to demonstrate the cybersecurity and software capabilities of Buildings Cybersecurity Framework (BCF) webtool. The webtool is designed based on BCF document and existing NIST standards. It’s capabilities and features are depicted through a building usecase with four different investment scenarios geared towards improving the cybersecurity posture of the building. BCF webtool also facilitates implementation of the goals outlined in Presidential Executive Order (EO) on Strengthening the Cybersecurity of Federal Networks and Critical Infrastructure (May 2017. In realization of the EO goals, BCF includes five core elements: Identify, Protect, Detect, Respond, and Recover, to help determine various policy and process level vulnerabilities and provide mitigation strategies. With the BCF webtool, an organization can perform a cybersecurity self-assessment; determine the current cybersecurity posture; define investment based goals to achieve a target state; connect the cybersecurity posture with business processes, functions, and continuity; and finally, develop plans to answer critical organizational cybersecurity questions. In this paper, the webtool and its core capabilities are depicted by performing an extensive comparative assessment over four different scenarios.

Scenarios for minor actinides transmutation in the framework of the French Act on Waste Management

International Nuclear Information System (INIS)

Coquelet-Pascal, C.; Meyer, M.; Tiphine, M.; Girieud, R.; Eschbach, R.; Chabert, C.; Garzenne, C.; Barbrault, P.; Van Den Durpel, L.; Caron-Charles, M.; Favet, D.; Arslan, M.; Caron-Charles, M.; Carlier, B.; Lefevre, J.C.

2013-01-01

In the framework of the French Act on Waste Management, options of minor actinides (MA) transmutation are studied, based on several scenarios of sodium fast reactor deployment. Basically, one of these scenarios considers the deployment of a 60 GWe SFR fleet in two steps (20 GWe from 2040 to 2050 and 40 GWe, as well as, from 2080 to 2100). For this scenario, the advantages and drawbacks of different transmutation options are evaluated: - transmutation of all minor actinides or only of americium; - transmutation in homogeneous mode (MA bearing fuel in all the core or just in the outer core) or in heterogeneous mode (MA bearing radial blankets). Scenarios have been optimised to limit the impacts of MA transmutation on the cycle: - reduction of the initial MA content in the core in the case of transmutation in homogeneous mode to reduce the impact on reactivity coefficients; - reduction of the number of rows of blankets and fuel decay heat in the case of transmutation in heterogeneous mode. The sensitivity of transmutation options to cycle parameters such as the fuel cooling time before transportation is also assessed. Thus, the transmutation of only americium in one row of radial blankets containing initially 10 pc % Am and irradiated during the same duration as the standard fuel assemblies appears to be a suitable solution to limit the transmutation impacts on fuel cycle and facilities. A comparison of results obtained with MA transmutation in dedicated systems is also presented with a symbiotic scenario considering ADS (accelerator-driven system) deployment to transmute MA together with a SFR fleet to produce energy. The MA inventory within the cycle is higher in the case of transmutation in ADS than in the case of transmutation in SFR. Considering the industrial feasibility of MA transmutation, it appears important to study 'independently' SFR deployment and MA transmutation. Consequently, scenarios of progressive introduction of MA options are assessed

HTR core physics analysis at NRG

International Nuclear Information System (INIS)

Kuijper, J.C.; Haas, J.B.M. de; Oppe, J.

2002-01-01

Since a number of years NRG is developing the HTR reactor physics code system PANTHERMIX. In PANTHERMIX the 3-D steady-state and transient core physics code PANTHER has been interfaced with the HTR thermal hydraulics code THERMIX to enable core follow and transient analyses on both pebble bed and block type HTR systems. Recently the capabilities of PANTHERMIX have been extended with the possibility to simulate the flow of pebbles through the core cavity and the (re)loading of pebbles on top of the core.The PANTHERMIX code system is being applied for the benchmark exercises for the Chinese HTR-10 and Japanese HTTR first criticality, calculating the critical loading, control rod worth and the isothermal temperature coefficients at zero power conditions. Also core physics calculations have been performed on an early version the South African PBMR design. The reactor physics properties of the reactor at equilibrium core loading have been studied as well as a selected run-in scenario, starting form fresh fuel. The recently developed reload option of PANTHERMIX was used extensively in these analyses. The examples shown demonstrate the capabilities of PANTHERMIX for performing steady-state and transient HTR core physics analyses. However, additional validation, especially for transient analyses, remains desirable. (author)

Using Mind Maps to Make Student Questioning Effective: Learning Outcomes of a Principle-Based Scenario for Teacher Guidance

Science.gov (United States)

Stokhof, Harry; de Vries, Bregje; Bastiaens, Theo; Martens, Rob

2018-01-01

Student questioning is an important learning strategy, but rare in many classrooms, because teachers have concerns if these questions contribute to attaining curricular objectives. Teachers face the challenge of making student questioning effective for learning the curriculum. To address this challenge, a principle-based scenario for guiding effective student questioning was developed and tested for its relevance and practicality in two previous studies. In the scenario, which consists of a sequence of pedagogical activities, mind maps support teachers and students to explore and elaborate upon a core curriculum, by raising, investigating, and exchanging student questions. In this paper, a follow-up study is presented that tested the effectiveness of the scenario on student outcomes in terms of attainment of curricular objectives. Ten teachers and their 231 students participated in the study. Pre- and posttest mind maps were used to measure individual and collective learning outcomes of student questioning. Findings show that a majority of students progressed in learning the core curriculum and elaborated upon it. The findings suggest that visualizing knowledge construction in a shared mind map supports students to learn a core curriculum and to refine their knowledge structures.

Test Specification of A1-1 Test for OECD-ATLAS Project

International Nuclear Information System (INIS)

Kang, Kyoung-Ho; Moon, Sang-Ki; Lee, Seung-Wook; Choi, Ki-Yong; Song, Chul-Hwa

2014-01-01

In the OECD-ATLAS project, design extension conditions (DECs) such as a station blackout (SBO) and a total loss of feed water (TLOFW) will be experimentally investigated to meet the international interests in the multiple high-risk DECs raised after the Fukushima accident. The proposed test matrix for the OECD-ATLAS project is summarized in Table 1.. In this study, detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG number 2 (SG-2). A SBO is one of the most important DECs in that without any proper operator actions, a total loss of heat sink leads to core uncover, to core damage, and ultimately a core melt-down scenario under high pressure. Due to this safety importance, a SBO is considered to be a base test item of the OECD-ATLAS project. A detailed specification of the first test named as A1-1 in the OECD-ATLAS project was described. The target scenario of the A1-1 test is a prolonged SBO with delayed supply of turbine-driven auxiliary feedwater to only SG-2 in order to consider an accident mitigation measure. The pre-test analysis using MARS code was performed with an aim of setting up the detailed test procedures for A1-1 test and also gaining the physical insights for a prolonged SBO transient. In the A1-1 test, a prolonged SBO transient will be simulated with two temporal phases: Phase (I) for conservative SBO transient without supply of turbine-driven auxiliary feedwater and Phase (II) for asymmetric cooling via single trained supply of turbine-driven auxiliary feedwater

A probabilistic analysis of rapid boron dilution scenarios

International Nuclear Information System (INIS)

Kohut, P.; Diamond, D.J.

1993-01-01

A probabilistic and deterministic analysis of a rapid boron dilution scenario related to reactor restart was performed. The event is initiated by a loss of off-site power during the startup dilution process. The automatic restart of the charging pump in such cases may lead to the accumulation of a diluted slug of water in the lower plenum. The restart of the reactor coolant pumps may send the diluted slug through the core, adding sufficient reactivity to overcome the shutdown margin and cause a power excursion. The concern is that the power excursion is sufficient in certain circumstances to cause fuel damage. The estimated core damage frequency based on the scoping analysis is 1.0--3.0E-05/yr for the plants analyzed. These are relatively significant values when compared to desirable goals. The analysis contained assumptions related to plant specific design characteristics which may lead to non-conservative estimates. The most important conservative assumptions were that mixing of the injected diluted water is insignificant and that fuel damage occurs when the slug passes through the core

The influence of core bypass flow during SBLOCA

International Nuclear Information System (INIS)

Maselj, A.; Jurkovic, M.

1996-01-01

Many parameters affect the behaviour of a NPP during a Small Break Loss of Coolant Accident (SBLOCA). The bypass flow between the core side and the downcomer is one of them. Different PWRs have different values of core bypass flow. In spite of the complexity of the real situation in the primary system during SBLOCA, some fundamental details of the phenomena can be explained with simplified mathematical models, which relate on basic parameters of the primary coolant. These models define the conditions for loop seal clearance and final results are confirmed with measured values. The analysis presented in the paper refers to Bethsy Test 9.1.b SB LOCA scenario, with variation of core bypass flow. Basic RELAP5 input model calculation results show very good agreement with the experimental data. The core liquid level depression before loop seal clearance is lower in case of smaller core bypass flow. This affects the fuel clad temperature because of different heat transfer mechanisms. Time of loop seal clearance is delayed with larger core bypass flow and consequently lower differential pressure between downcomer and core. (author)

Modelling guidelines for core exit temperature simulations with system codes

Energy Technology Data Exchange (ETDEWEB)

Freixa, J., E-mail: jordi.freixa-terradas@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Martínez-Quiroga, V., E-mail: victor.martinez@nortuen.com [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain); Zerkak, O., E-mail: omar.zerkak@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen (Switzerland); Reventós, F., E-mail: francesc.reventos@upc.edu [Department of Physics and Nuclear Engineering, Technical University of Catalonia (UPC) (Spain)

2015-05-15

Highlights: • Core exit temperature is used in PWRs as an indication of core heat up. • Modelling guidelines of CET response with system codes. • Modelling of heat transfer processes in the core and UP regions. - Abstract: Core exit temperature (CET) measurements play an important role in the sequence of actions under accidental conditions in pressurized water reactors (PWR). Given the difficulties in placing measurements in the core region, CET readings are used as criterion for the initiation of accident management (AM) procedures because they can indicate a core heat up scenario. However, the CET responses have some limitation in detecting inadequate core cooling and core uncovery simply because the measurement is not placed inside the core. Therefore, it is of main importance in the field of nuclear safety for PWR power plants to assess the capabilities of system codes for simulating the relation between the CET and the peak cladding temperature (PCT). The work presented in this paper intends to address this open question by making use of experimental work at integral test facilities (ITF) where experiments related to the evolution of the CET and the PCT during transient conditions have been carried out. In particular, simulations of two experiments performed at the ROSA/LSTF and PKL facilities are presented. The two experiments are part of a counterpart exercise between the OECD/NEA ROSA-2 and OECD/NEA PKL-2 projects. The simulations are used to derive guidelines in how to correctly reproduce the CET response during a core heat up scenario. Three aspects have been identified to be of main importance: (1) the need for a 3-dimensional representation of the core and Upper Plenum (UP) regions in order to model the heterogeneity of the power zones and axial areas, (2) the detailed representation of the active and passive heat structures, and (3) the use of simulated thermocouples instead of steam temperatures to represent the CET readings.

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

International Nuclear Information System (INIS)

Ott, L.J.; Hagen, S.

1993-01-01

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

Analysis of natural convection in volumetrically-heated melt pools

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.

1996-12-01

Results of series of studies on natural convection heat transfer in decay-heated core melt pools which form in a reactor lower plenum during the progression of a core meltdown accident are described. The emphasis is on modelling and prediction of turbulent heat transfer characteristics of natural convection in a liquid pool with an internal energy source. Methods of computational fluid dynamics, including direct numerical simulation, were applied for investigation

Testing the Formation Scenarios of Binary Neutron Star Systems with Measurements of the Neutron Star Moment of Inertia

Science.gov (United States)

Newton, William G.; Steiner, Andrew W.; Yagi, Kent

2018-03-01

Two low-mass (M slope of the nuclear symmetry energy L. We find that, if either J0737-3039B or the J1756-2251 companion were formed in a US-SN, no more than 0.06 M ⊙ could have been lost from the progenitor core. Furthermore, a measurement of the moment of inertia of J0737-3039A to within 10% accuracy can discriminate between formation scenarios and, given current constraints on the predicted core mass loss, potentially rule them out. Advanced LIGO can potentially measure the neutron star tidal polarizability to equivalent accuracy which, using the I-Love-Q relations, would obtain similar constraints on the formation scenarios. Such information would help constrain important aspects of binary evolution used for population synthesis predictions of the rate of binary neutron star mergers and resulting electromagnetic and gravitational wave signals. Further progress needs to be made in modeling the core-collapse process that leads to low-mass neutron stars, particularly in making robust predictions for the mass loss from the progenitor core.

Neutron dynamics of fast-spectrum dedicated cores for waste transmutation

International Nuclear Information System (INIS)

Massara, S.

2002-04-01

Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

Intercode comparison of SBO scenario for AHWR

International Nuclear Information System (INIS)

Srivastava, A.; Kumar, Rajesh; Chatterjee, B.; Vijayan, P.K.

2015-01-01

This paper outlines the assessment of station blackout scenario for AHWR using last version of the French best estimate computer code CATHARE2/V2.5 2 and its comparison with RELAP5/mod3.2 findings. First, it explains the modelling of main heat transport system of AHWR and isolation Condenser loop along with GDWP in CATHARE2 followed by thermal hydraulic safety assessment of station blackout scenario and comparison of predictions with RELAP5 findings. The proposed Advanced Heavy Water Reactor is a 920 MWth Thorium based vertical pressure tube type boiling light water cooled and heavy water moderated reactor. One of the important passive design features of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all allowed power levels with no primary coolant pumps.The prolonged SBO has been analyzed for AHWR using best estimate code CATHARE and results are compared with already available results of RELAP5 code. The ICs are able to remove decay heat with the help of GDWP and maintain core temperatures well with-in the limit. Decay heat is removed passively by GDWP initially by sensible heating and later by boil off as seen in long term calculation done with RELAP5. It has been found in that analysis, IC system is capable to remove decay heat for more than 7 days. The decay heat removal through IC path along with passive moderator and end shield cooling keeps the integrity of different system and maintains the core temperature well below the acceptance limit

Scenario Planning Implementation : A Case of a Malaysian Organisation

OpenAIRE

Lim, Lee Chin

2007-01-01

In this dynamic and ever changing era, uncertainties are inevitable. A successful organisation is one that is able to address the uncertainties in a world that is complex and unpredictable. Scenario planning is a strategic management tool that offers an alternative approach to planning in a world where linear logic does not prevail (Geons, 2001). It is a method for learning about the future by understanding the uncertainties and driving forces affecting an organisation’s future. Its core bene...

Prospects for future uranium savings through LWRs with high performance cores

International Nuclear Information System (INIS)

Mochida, T.; Yamamoto, T.; Sasaki, M.; Matsuura, H.; Ueji, M.; Murata, T.; Kanda, K.; Oka, Y.; Kondo, S.

1995-01-01

Since 1986, Nuclear Power Engineering Cooperation (NUPEC) has been studying four types of LWR high performance core concepts (i.e., the uranium saving core I (USC-I), the uranium saving core II (USC-II), the high moderation core (HMC) and the low moderation core (LMC)), which aim at improvement of uranium and plutonium utilization. After the evaluation of fundamental core performance and uranium and plutonium material balance for each reactor, potential uranium savings with different reactor strategies are evaluated for the Japanese scenario with assumption of the growth of future nuclear power plant generation, annual reprocessing capacity and schedules for the introduction of high performance core. At 2030, about 3-6% savings in uranium demand are expected by USC-I or USC-II strategy, while about 14% savings by HMC strategy and about 8% by LMC strategy. (author)

Safety oriented LWR research. Annual report 1990

International Nuclear Information System (INIS)

1991-07-01

The contributions describe phenomenons of severe fuel damage and aspects of core meltdown accidents. These accidents deal with aerosol behaviour and ventilation systems and the methods for assessing and reducing the radiological concequences of nuclear accidents. Other contributions describe selected questions of safety of HCLWR type reactors. (DG)

The possibility and the effects of a steam explosion in the BWR lower head on recriticality of a BWR core

International Nuclear Information System (INIS)

Sehgal, B.R.; Dinh, T.N.

2002-12-01

The report describes an analysis considering a BWR postulated severe accident scenario during which the late vessel automatic depressurization brings the water below the level of the bottom core plate. The subsequent lack of ECCS leads to core heat up during which the control rods melt and the melt deposits on the core plate. At that point of time in the scenario, the core fuel bundles are still intact and the Zircaloy clad oxidation is about to start. The objective of the study is to provide the conditions of reflood into the hot core due to the level swell or a slug delivered from the lower head as the control rod melt drops into the water. These conditions are employed in the neutronic analysis with the RECRIT code to determine if the core recriticality may be achieved. (au)

A 3.55 keV line from DM →a→γ: predictions for cool-core and non-cool-core clusters

Energy Technology Data Exchange (ETDEWEB)

Conlon, Joseph P.; Powell, Andrew J. [Rudolf Peierls Centre for Theoretical Physics, University of Oxford, 1 Keble Road, Oxford, OX1 3NP (United Kingdom)

2015-01-13

We further study a scenario in which a 3.55 keV X-ray line arises from decay of dark matter to an axion-like particle (ALP), that subsequently converts to a photon in astrophysical magnetic fields. We perform numerical simulations of Gaussian random magnetic fields with radial scaling of the magnetic field magnitude with the electron density, for both cool-core 'Perseus' and non-cool-core 'Coma' electron density profiles. Using these, we quantitatively study the resulting signal strength and morphology for cool-core and non-cool-core clusters. Our study includes the effects of fields of view that cover only the central part of the cluster, the effects of offset pointings on the radial decline of signal strength and the effects of dividing clusters into annuli. We find good agreement with current data and make predictions for future analyses and observations.

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

Energy Technology Data Exchange (ETDEWEB)

Brown, C.S., E-mail: csbrown3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Raleigh, NC 27695-7909 (United States); Zhang, H., E-mail: Hongbin.Zhang@inl.gov [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3870 (United States); Kucukboyaci, V., E-mail: kucukbvn@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States); Sung, Y., E-mail: sungy@westinghouse.com [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2016-12-01

Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

Best estimate plus uncertainty analysis of departure from nucleate boiling limiting case with CASL core simulator VERA-CS in response to PWR main steam line break event

International Nuclear Information System (INIS)

Brown, C.S.; Zhang, H.; Kucukboyaci, V.; Sung, Y.

2016-01-01

Highlights: • Best estimate plus uncertainty (BEPU) analyses of PWR core responses under main steam line break (MSLB) accident. • CASL’s coupled neutron transport/subchannel code VERA-CS. • Wilks’ nonparametric statistical method. • MDNBR 95/95 tolerance limit. - Abstract: VERA-CS (Virtual Environment for Reactor Applications, Core Simulator) is a coupled neutron transport and thermal-hydraulics subchannel code under development by the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS was applied to simulate core behavior of a typical Westinghouse-designed 4-loop pressurized water reactor (PWR) with 17 × 17 fuel assemblies in response to two main steam line break (MSLB) accident scenarios initiated at hot zero power (HZP) at the end of the first fuel cycle with the most reactive rod cluster control assembly stuck out of the core. The reactor core boundary conditions at the most DNB limiting time step were determined by a system analysis code. The core inlet flow and temperature distributions were obtained from computational fluid dynamics (CFD) simulations. The two MSLB scenarios consisted of the high and low flow situations, where reactor coolant pumps either continue to operate with offsite power or do not continue to operate since offsite power is unavailable. The best estimate plus uncertainty (BEPU) analysis method was applied using Wilks’ nonparametric statistical approach. In this demonstration of BEPU application, 59 full core simulations were performed for each accident scenario to provide the minimum departure from nucleate boiling ratio (MDNBR) at the 95/95 (95% probability with 95% confidence level) tolerance limit. A parametric goodness-of-fit approach was also applied to the results to obtain the MDNBR value at the 95/95 tolerance limit. Initial sensitivity analysis was performed with the 59 cases per accident scenario by use of Pearson correlation coefficients. The results show that this typical PWR core

200 Gbit/s 16QAM WDM transmission over a fully integrated cladding pumped 7-Core MCF System

DEFF Research Database (Denmark)

Castro, C.; Jain, S.; Jung, Y.

2017-01-01

A complete, realistic integrated system is investigated, consisting of directly spliced 7-core MCF, cladding-pumped 7-core amplifiers, isolators, and couplers. The system is demonstrated in a 16QAM C-band WDM scenario over 720 km....

Biomass Scenario Model Scenario Library: Definitions, Construction, and Description

Energy Technology Data Exchange (ETDEWEB)

Inman, D.; Vimmerstedt, L.; Bush, B.; Peterson, S.

2014-04-01

Understanding the development of the biofuels industry in the United States is important to policymakers and industry. The Biomass Scenario Model (BSM) is a system dynamics model of the biomass-to-biofuels system that can be used to explore policy effects on biofuels development. Because of the complexity of the model, as well as the wide range of possible future conditions that affect biofuels industry development, we have not developed a single reference case but instead developed a set of specific scenarios that provide various contexts for our analyses. The purpose of this report is to describe the scenarios that comprise the BSM scenario library. At present, we have the following policy-focused scenarios in our library: minimal policies, ethanol-focused policies, equal access to policies, output-focused policies, technological diversity focused, and the point-of-production- focused. This report describes each scenario, its policy settings, and general insights gained through use of the scenarios in analytic studies.

Severe fuel damage in steam and helium environments observed in in-reactor experiments

International Nuclear Information System (INIS)

Saito, S.; Shiozawa, S.

1984-01-01

The bahavior of severe fuel damages has been studied in gaseous environments simulating core uncovery accidents in the in-reactor experiments utilizing the NSRR. Two types of cladding relocation modes, azimuthal flow and melt-down, were revealed through the parametric experiments. The azimuthal flow was evident in an oxidizing environment in case of no oxide film break. The melt-down can be categorized into flow-down and move-down, according to the velocity of the melt-down. Cinematographies showed that the flow-down was very fast as water flows down while the move-down appeared to be much slower. The flow-down was possible in an unoxidizing environment, whereas the move-down of molten cladding occured through a crack induced in an oxide film in an oxidizing environment. The criterion of the relocation modes was developed as a function of peak cladding temperature and oxidation condition. It was also found that neither immediate quench nor fuel fracture occurred upon flooding when cladding temperature was about 1800 0 C at water injection. The external mechanical force is needed for fuel fracture. (orig.)

Standards and guidelines should be rules between licensees and regulators

International Nuclear Information System (INIS)

Narabayashi, Tadashi

2014-01-01

The 2011 off the Pacific coast of Tohoku Pacific Earthquake and the Tsunami gave the serious damage to the Fukushima-Daiichi Nuclear Power Plants (NPPs). The accidents occurred in Unit 1, 2, 3 and 4. It is said that the height of tsunami attacked Fukushima NPP was more than 14m. After 50 minutes from the automatic shut-down, tsunami attacked the NPPs in Fukushima Daiichi NPPs. For example, the Unit 1 lost A/C power caused the loss of water injection function; it made the core meltdown and unusual increase of PCV pressure in the midnight of March 11th to 12th morning. Though the Unit one has the Isolation Condenser Core Cooling system, it was stopped by the operator to keep the cooling rate of 55degC/h. Finally, the isolation signal was transmitted from the control room to the motor driven isolation valves when the control room's battery discharged. It was the initiation of the core meltdown. The lessons from the accidents, we should improve the nuclear safety regulation through the innovation of regulatory rules and safety standards. Standards and guidelines should be rules between licensees and regulators. (author)

Top-down freezing in a Fe-FeS core and Ganymede's present-day magnetic field

Science.gov (United States)

Rückriemen, Tina; Breuer, Doris; Spohn, Tilman

2018-06-01

Ganymede's core most likely possesses an active dynamo today, which produces a magnetic field at the surface of ∼ 719 nT. Thermochemical convection triggered by cooling of the core is a feasible power source for the dynamo. Experiments of different research groups indicate low pressure gradients of the melting temperatures for Fe-FeS core alloys at pressures prevailing in Ganymede's core ( < 10 GPa). This may entail that the core crystallizes from the top instead of from the bottom as is expected for Earth's core. Depending on the core sulfur concentration being more iron- or more sulfur-rich than the eutectic concentration either snowing iron crystals or a solid FeS layer can form at the top of the core. We investigate whether these two core crystallization scenarios are capable of explaining Ganymede's present magnetic activity. To do so, we set up a parametrized one-dimensional thermal evolution model. We explore a wide range of parameters by running a large set of Monte Carlo simulations. Both freezing scenarios can explain Ganymede's present-day magnetic field. Dynamos of iron snow models are rather young ( < 1 Gyr), whereas dynamos below the FeS layer can be both young and much older ( ∼ 3.8 Gyr). Successful models preferably contain less radiogenic heat sources in the mantle than the chondritic abundance and show a correlation between the reference viscosity in the mantle and the initial core sulfur concentration.

Analysis of natural convection in volumetrically-heated melt pools

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-01

Results of series of studies on natural convection heat transfer in decay-heated core melt pools which form in a reactor lower plenum during the progression of a core meltdown accident are described. The emphasis is on modelling and prediction of turbulent heat transfer characteristics of natural convection in a liquid pool with an internal energy source. Methods of computational fluid dynamics, including direct numerical simulation, were applied for investigation. Refs, figs, tabs.

Radioactive contamination of Danish territory after core-melt accidents at the Barsebaeck power plant

International Nuclear Information System (INIS)

Gjoerup, H.L.; Jensen, N.O.; Hedemann Jensen, P.; Kristensen, L.; Nielsen, O.J.; Petersen, E.L.; Petersen, T.; Roed, J.; Thykier-Nielsen, S.; Heikel Vinter, F.; Warming, L.; Aarkrog, A.

1982-03-01

An assessment is made of the radioactive contamination of Danish territory in the event of a core-melt accident at the Barsebaeck nuclear power plant in Sweden. Accidents including both core melt-down and containment failure are considered. Consequences are calculated for a BWR-3 release under common meteorological conditions and for a BWR-2 release under extreme meteorological conditions. Calculations are based on experiments and theoretical work relating to deposition velocities for different types of surface, shielding effect of structures, and weathering. The effects are described of different dose-reducing measures, e.g., decontamination, relocation, destruction of contaminated foodstuffs. The collective effective dose equivalent from external gamma radiation from deposited activity integrated over a time period of 30 years, is calculated to be 3.6 Megamanrem in the BWR-3 case without dose-reducing measures. For the BWR-2 case, the corresponding dose is approx. 41 Megamanrem. A combination of temporary relocation, hosing of roads etc. and digging of gardens is estimated to reduce these doses to approx. 2.5 Megamanrem and approx. 15 Megamanrem, respectively. The collective committed effective dose equivalent from the consumption of contaminated foodstuffs is calculated to 23 Megamanrem in the BWR-3 case without dose-reducing measures. This dose could be reduced to 0.2 Megamanrem if contaminated crops are destroyed during the first year after the accident and if changes are made in agricultural production in the contaminated area. The corresponding doses in the BWR-2 case would be 197 Megamanrem and 1.4 Megmanrem, respectively. (author)

Modeling and analysis of selected organization for economic cooperation and development PKL-3 station blackout experiments using TRACE

Directory of Open Access Journals (Sweden)

Roman Mukin

2018-04-01

Full Text Available A series of tests dedicated to station blackout (SBO accident scenarios have been recently performed at the Primärkreislauf-Versuchsanlage (primary coolant loop test facility; PKL facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal–hydraulic code TRACE (v5.0 Patch4 of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for

Scenario planning.

Science.gov (United States)

Enzmann, Dieter R; Beauchamp, Norman J; Norbash, Alexander

2011-03-01

In facing future developments in health care, scenario planning offers a complementary approach to traditional strategic planning. Whereas traditional strategic planning typically consists of predicting the future at a single point on a chosen time horizon and mapping the preferred plans to address such a future, scenario planning creates stories about multiple likely potential futures on a given time horizon and maps the preferred plans to address the multiple described potential futures. Each scenario is purposefully different and specifically not a consensus worst-case, average, or best-case forecast; nor is scenario planning a process in probabilistic prediction. Scenario planning focuses on high-impact, uncertain driving forces that in the authors' example affect the field of radiology. Uncertainty is the key concept as these forces are mapped onto axes of uncertainty, the poles of which have opposed effects on radiology. One chosen axis was "market focus," with poles of centralized health care (government control) vs a decentralized private market. Another axis was "radiology's business model," with one pole being a unified, single specialty vs a splintered, disaggregated subspecialty. The third axis was "technology and science," with one pole representing technology enabling to radiology vs technology threatening to radiology. Selected poles of these axes were then combined to create 3 scenarios. One scenario, termed "entrepreneurialism," consisted of a decentralized private market, a disaggregated business model, and threatening technology and science. A second scenario, termed "socialized medicine," had a centralized market focus, a unified specialty business model, and enabling technology and science. A third scenario, termed "freefall," had a centralized market focus, a disaggregated business model, and threatening technology and science. These scenarios provide a range of futures that ultimately allow the identification of defined "signposts" that can

Adaptive scenarios: a training model for today's public health workforce.

Science.gov (United States)

Uden-Holman, Tanya; Bedet, Jennifer; Walkner, Laurie; Abd-Hamid, Nor Hashidah

2014-01-01

With the current economic climate, money for training is scarce. In addition, time is a major barrier to participation in trainings. To meet the public health workforce's rising demand for training, while struggling with less time and fewer resources, the Upper Midwest Preparedness and Emergency Response Learning Center has developed a model of online training that provides the public health workforce with individually customized, needs-based training experiences. Adaptive scenarios are rooted in case-based reasoning, a learning approach that focuses on the specific knowledge needed to solve a problem. Proponents of case-based reasoning argue that learners benefit from being able to remember previous similar situations and reusing information and knowledge from that situation. Adaptive scenarios based on true-to-life job performance provide an opportunity to assess skills by presenting the user with choices to make in a problem-solving context. A team approach was used to develop the adaptive scenarios. Storylines were developed that incorporated situations aligning with the knowledge, skills, and attitudes outlined in the Public Health Preparedness and Response Core Competency Model. This article examines 2 adaptive scenarios: "Ready or Not? A Family Preparedness Scenario" and "Responding to a Crisis: Managing Emotions and Stress Scenario." The scenarios are available on Upper Midwest Preparedness and Emergency Response Learning Center's Learning Management System, the Training Source (http://training-source.org). Evaluation data indicate that users' experiences have been positive. Integrating the assessment and training elements of the scenarios so that the training experience is uniquely adaptive to each user is one of the most efficient ways to provide training. The opportunity to provide individualized, needs-based training without having to administer separate assessments has the potential to save time and resources. These adaptive scenarios continue to be

Sensitivity of Transmutation Capability to Recycling Scenarios in KALIMER-600 TRU Burner

International Nuclear Information System (INIS)

Lee, Yong Kyo; Kim, Myung Hyun

2013-01-01

The purpose of this study is to test transmutation and design feasibility of KALIMER burner caused from many limitations in recycling options; such as low recovery factors and external feed. Design impact from many recycling options will be tested as a sensitivity to various recycling process parameters under many recycling scenarios. Through this study, possibilities when Pyro-processing is realized with SFR can be expected in the recycling scenarios. For the development of sodium-cooled fast reactor(SFR) technology, prototype KALIMER plant is now under R and D stage in Korea. For the future application of SFR for waste transmutation, KALIMER core was designed for TRU burner by KAERI. Feasibility of TRU burner cannot be evaluated exactly because overall functional parameters in pyro-processing recycling process has not been verified yet. There is great possibility to accept undesirable process functions in pyro-processing. Only TRU nuclides composition a little differs between PWR SF and CANDU SF so first scenario has no problem operating SFR. In second scenario, the radiotoxicity of waste at 99% of TRU RF have to be confirmed whether it is proper level to reposit as Low and Intermediate Level Wastes or not. And the reactor safety at high RF of RE must be inspected. Not only third scenario but also several scenarios for good measure are being calculated and will be evaluated

Use of decision trees for evaluating severe accident management strategies in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclerar Engineering; Lee, Yongjin; Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of). School of Energy Systems Engineering

2016-07-15

Accident management strategies are defined to innovative actions taken by plant operators to prevent core damage or to maintain the sound containment integrity. Such actions minimize the chance of offsite radioactive substance leaks that lead to and intensify core damage under power plant accident conditions. Accident management extends the concept of Defense in Depth against core meltdown accidents. In pressurized water reactors, emergency operating procedures are performed to extend the core cooling time. The effectiveness of Severe Accident Management Guidance (SAMG) became an important issue. Severe accident management strategies are evaluated with a methodology utilizing the decision tree technique.

Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

2013-06-01

Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

Management strategies in hospitals: scenario planning

Directory of Open Access Journals (Sweden)

Ghanem, Mohamed

2015-06-01

Full Text Available Background: Instead of waiting for challenges to confront hospital management, doctors and managers should act in advance to optimize and sustain value-based health. This work highlights the importance of scenario planning in hospitals, proposes an elaborated definition of the stakeholders of a hospital and defines the influence factors to which hospitals are exposed to. Methodology: Based on literature analysis as well as on personal interviews with stakeholders we propose an elaborated definition of stakeholders and designed a questionnaire that integrated the following influence factors, which have relevant impact on hospital management: political/legal, economic, social, technological and environmental forces. These influence factors are examined to develop the so-called critical uncertainties. Thorough identification of uncertainties was based on a “Stakeholder Feedback”. Results: Two key uncertainties were identified and considered in this study: According to the developed scenarios, complementary education of the medical staff as well as of non-medical top executives and managers of hospitals was the recommended core strategy. Complementary scenario-specific strategic options should be considered whenever needed to optimize dealing with a specific future development of the health care environment. Conclusion: Strategic planning in hospitals is essential to ensure sustainable success. It considers multiple situations and integrates internal and external insights and perspectives in addition to identifying weak signals and “blind spots”. This flows into a sound planning for multiple strategic options. It is a state of the art tool that allows dealing with the increasing challenges facing hospital management.

Management strategies in hospitals: scenario planning.

Science.gov (United States)

Ghanem, Mohamed; Schnoor, Jörg; Heyde, Christoph-Eckhard; Kuwatsch, Sandra; Bohn, Marco; Josten, Christoph

2015-01-01

Instead of waiting for challenges to confront hospital management, doctors and managers should act in advance to optimize and sustain value-based health. This work highlights the importance of scenario planning in hospitals, proposes an elaborated definition of the stakeholders of a hospital and defines the influence factors to which hospitals are exposed to. Based on literature analysis as well as on personal interviews with stakeholders we propose an elaborated definition of stakeholders and designed a questionnaire that integrated the following influence factors, which have relevant impact on hospital management: political/legal, economic, social, technological and environmental forces. These influence factors are examined to develop the so-called critical uncertainties. Thorough identification of uncertainties was based on a "Stakeholder Feedback". Two key uncertainties were identified and considered in this study: the development of workload for the medical staff the profit oriented performance of the medical staff. According to the developed scenarios, complementary education of the medical staff as well as of non-medical top executives and managers of hospitals was the recommended core strategy. Complementary scenario-specific strategic options should be considered whenever needed to optimize dealing with a specific future development of the health care environment. Strategic planning in hospitals is essential to ensure sustainable success. It considers multiple situations and integrates internal and external insights and perspectives in addition to identifying weak signals and "blind spots". This flows into a sound planning for multiple strategic options. It is a state of the art tool that allows dealing with the increasing challenges facing hospital management.

Innovative TRU Burners and Fuel Cycles Options for Phase-Out and Regional Scenarios

International Nuclear Information System (INIS)

Vezzoni, B.; Gabrielli, F.; Rineiski, A.; Schwenk-Ferrero, A.; Andriolo, L.; Maschek, W.

2015-01-01

Partitioning and transmutation (P and T) technologies may be considered either for minor actinides (MAs) inventory stabilisation (typical for on-going/regional scenarios) or for a drastic reduction of the transuranics inventory (as in phasing-out scenarios). In this paper, two sodium-cooled fast reactor cores, based on the French ASTRID design and characterised by different amounts of MAs in the fuel, are proposed. Attention focuses on the safety and on the burning performances of the systems. The behaviour of the systems under dynamic conditions has been investigated considering phasing-out and on-going fuel cycle scenarios. The results demonstrate the flexibility of such systems when employed in different kinds of fuel cycles. The impact of different parameters, such as the initial isotopic vector (and Cm content) and the cooling time before reprocessing, on the simulation results is investigated as well. (authors)

Hydrogen-Poor Core-Collapse Supernovae

Science.gov (United States)

Pian, Elena; Mazzali, Paolo A.

Hydrogen-poor core-collapse supernovae (SNe) signal the explosive death of stars more massive than the progenitors of hydrogen-rich core-collapse supernovae, i.e., approximately in the range 15-50 M⊙ in main sequence. Since hydrogen-poor core-collapse supernovae include those that accompany gamma-ray bursts (GRBs), which were all rigorously identified with type Ic supernovae, their explosion energies cover almost two decades. The light curves and spectra are consequently very heterogeneous and often bear the signature of an asymmetric, i.e., aspherical, explosion. Asphericity is best traced by early-time (within days of the explosion) optical spectropolarimetry and by late-epoch (more than ˜ 100 days after explosion) low-resolution spectroscopy. While the relationship between hydrogen-poor core-collapse supernovae to hydrogen-poor super-luminous supernovae is not understood, a known case of association between an ultra-long gamma-ray burst and a very luminous hydrogen-poor supernova may help unraveling the connection. This is tantalizingly pointing to a magnetar powering source for both phenomena, although this scenario is still highly speculative. Host galaxies of hydrogen-poor supernovae are always star forming; in those of completely stripped supernovae and gamma-ray burst supernovae, the spatial distribution of the explosions follows the blue/ultraviolet light, with a correlation that is more than linear.

European community light water reactor safety research projects. Experimental issue

International Nuclear Information System (INIS)

1975-01-01

Research programs on light water reactor safety currently carried out in the European Community are presented. They cover: accident conditions (LOCA, ECCS, core meltdown, external influences, etc...), fault and accident prevention and means of mitigation, normal operation conditions, on and off site implications and equipment under severe accident conditions, and miscellaneous subjects

Study of diluting and absorber materials to control reactivity during a postulated core melt down accident in Generation IV reactors

International Nuclear Information System (INIS)

Plevacova, K.

2010-01-01

In order to limit the consequences of a hypothetical core meltdown accident in Generation IV Sodium Fast Reactors, absorber materials in or near the core, such as boron carbide B 4 C, and diluting materials in the core catcher will be used to prevent recriticality within the mixture of molten oxide fuel and molten structures called corium. The aim of the PhD thesis was to select materials of both types and to understand their behaviour during their interaction with corium, from chemical and thermodynamic point of view. Concerning B 4 C, thermodynamic calculations and experiments agree with the formation of two immiscible phases at high temperature in the B 4 C - UO 2 system: one oxide and one boride. This separation of phases can reduce the efficiency of the neutrons absorption inside the molten fuel contained in the oxide phase. Moreover, a volatilization of a part of the boron element can occur. According to these results, the necessary quantity of B 4 C to be introduced should be reconsidered for postulated severe accident sequence. Other solution could be the use of Eu 2 O 3 or HfO 2 as absorber material. These oxides form a solid solution with the oxide fuel. Concerning the diluting materials, mixed oxides Al 2 O 3 - HfO 2 and Al 2 O 3 - Eu 2 O 3 were preselected. These systems being completely unknown to date at high temperature in association with UO 2 , first points on the corresponding ternary phase diagrams were researched. Contrary to Al 2 O 3 - Eu 2 O 3 - UO 2 system, the Al 2 O 3 - HfO 2 - UO 2 mixture presents only one eutectic and thus only one solidification path which makes easier forecasting the behaviour of corium in the core catcher. (author) [fr

Attributes Of Quality Scenarios/Scenario Sets Used In Software Requirements Elicitation

National Research Council Canada - National Science Library

Braun, Kimberly

1997-01-01

.... This thesis examines scenarios used in software requirements elicitation. Many different definitions, formats, and ideas exist on scenarios, but no thorough work has been done on what makes a good, quality scenario and scenario set...

Evaluation of long-term post-accident core cooling of Three Mile Island Unit 2

Energy Technology Data Exchange (ETDEWEB)

None

1979-04-15

On the basis of current understanding of the accident scenario and available data, the staff reports here on its evaluation of the condition of the core and the core flow resistance as it might affect ability to cool the core by natural circulation. The natural circulation cooling capability of TMI-2 for the estimated core flow resistance and a variety of other conditions is evaluated and a comparison of the Base Case and off-nominal plant configurations is presented. The potential for and effects of natural convection core cooling are addressed, and the staff recommendations for reactor performance acceptance criteria upon initiation of natural convection are presented.

Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs

Energy Technology Data Exchange (ETDEWEB)

Chun, Ji-Han, E-mail: chunjh@kaeri.re.kr; Lim, Sung-Won; Chung, Bub-Dong; Lee, Won-Jae

2015-08-15

Highlights: • Thermal conductivity model of the FCM fuel was developed and adopted in the MARS. • Scoping analysis for candidate FCM FAs was performed to select feasible FA. • Preliminary safety criteria for FCM fuel and SiC/Zr cladding were set up. • Enhanced safety margin and accident tolerance for FCM-SiC/Zr core were demonstrated. - Abstract: The FCM fueled cores proposed as an accident tolerant concept is assessed against the design-basis-accident (DBA) and the beyond-DBA (BDBA) scenarios using MARS code. A thermal conductivity model of FCM fuel is incorporated in the MARS code to take into account the effects of irradiation and temperature that was recently measured by ORNL. Preliminary analyses regarding the initial stored energy and accident tolerant performance were carried out for the scoping of various cladding material candidates. A 16 × 16 FA with SiC-coated Zircalloy cladding was selected as the feasible conceptual design through a preliminary scoping analysis. For a selected design, safety analyses for DBA and BDBA scenarios were performed to demonstrate the accident tolerance of the FCM fueled core. A loss of flow accident (LOFA) scenario was selected for a departure-from-nucleate-boiling (DNB) evaluation, and large-break loss of coolant accident (LBLOCA) scenario for peak cladding temperature (PCT) margin evaluation. A control element assembly (CEA) ejection accident scenario was selected for peak fuel enthalpy and temperature. Moreover, a station blackout (SBO) and LBLOCA without a safety injection (SI) scenario were selected as a BDBA. It was demonstrated that the DBA safety margin of the FCM core is satisfied and the time for operator actions for BDBA s is evaluated.

Early Dynamics of the Moon's Core

Science.gov (United States)

Cuk, Matija; Hamilton, Douglas; Stewart, Sarah T.

2018-04-01

The Moon has a small molten iron core (Williams et al. 2006). Remanent magnetization in lunar rocks likely derives from a past lunar dynamo (Wieczorek 2018 and references therein), which may have been powered by differential precession between the mantle and the core. The rotations of the lunar mantle and core were largely decoupled for much of lunar history, with a large mutual offset during the Cassini State Transition (Meyer and Wisdom, 2011). It is likely that the past work underestimated lunar obliquities, and therefore core offsets, during early lunar history (Cuk et al. 2016). Here we investigate the dynamics of the lunar core and mantle using a Lie-Poisson numerical integrator (Touma and Wisdom 2001) which includes interactions between triaxial core and mantle, as well as all gravitational and tidal effects included in the model of Cuk et al. (2016). Since we assume a rigid triaxial mantle, this model is applicable to the Moon only once it has acquired its current shape, which probably happened before the Moon reached 25 Earth radii. While some details of the core dynamics depend on our assumptions about the shape of the lunar core-mantle boundary, we can report some robust preliminary findings. The presence of the core does not change significantly the evolutionary scenario of Cuk et al. (2016). The core and mantle are indeed decoupled, with the core having a much smaller obliquity to the ecliptic than the mantle for almost all of the lunar history. The core was largely in an equivalent of Cassini State 2, with the vernal equinoxes (wrt the ecliptic) of the core and the mantle being anti-aligned. The core-mantle spin axis offset has been very large during the Moon's first billion years (this is true both in canonical and high-inclination tidal evolution), causing the lunar core to be sub-synchronous. If the ancient lunar magnetic dipole was rotating around the core axis that was inclined to the Moon's spin axis, then the magnetic poles would move across

Core concept of fast power reactor with zero sodium void reactivity

International Nuclear Information System (INIS)

Matveev, V.I.; Chebeskov, A.N.; Krivitsky, I.Y.

1991-01-01

The paper presents a core concept of BN-800 - type fast power reactor with zero sodium void reactivity (SVR). Consideration is given to the layout-and some design features of such a core. Some considerations on the determination of the required SVR value as one of the fast reactor safety criteria in accidents with coolant boiling are presented. Some methodical considerations an the development of calculation models that give a correct description of the new core features are stated. The results of the integral SVR calculation studies are included. reactivity excursions under different scenarios of sodium boiling are estimated, some corrections into the calculated SVR value are discussed. (author)

Preventing another Chernobyl: codes, practices, and the role of new technology

International Nuclear Information System (INIS)

Egan, J.R.

1988-01-01

Preventative steps to prevent and mitigate the consequences of a nuclear accident are considered. The national and international legislation now available is mentioned elsewhere in the book. This chapter discusses the role of codes of conduct and personal agreements which could also prevent nuclear accidents. It is suggested that a radical departure from existing nuclear technology is needed. In particular some definite goals and objectives are suggested. The first is that there should be no meltdowns - all reactors must be physically incapable of core meltdown or explosion under any circumstances. This would mean no huge emergency evacuation plans, no need for expensive emergency core cooling systems. No spiralling cost overruns, no nuclear waste problem, no more huge reactors too big for national electrical systems, no military complications, no intractable decommissionings and no quality control problems are also given as goals. Some of these goals could be achieved with inherently safe reactor designs, much smaller reactors and factory assembly, rather than on-site construction. Other suggestions are also made to achieve the proposed new nuclear technology. (U.K.)

Lessons learned from first generation nuclear plant probabalistic risk assessments

International Nuclear Information System (INIS)

Garrick, B.J.

1984-01-01

The paper by Garrick summarizes the state-of-the-art in what are perhaps the most archetypical probabilistic risk assessments (PRAs). Because of its unique regulatory environment and because of the high levels of perceived (not necessarily actual) risk, the nuclear industry more than any other has been concerned with quantitative risk analysis. Garrick's paper summarizes the lessons learned from ten PRA's conducted in the nuclear industry, including six that can be characterized as full-scope risk studies. Most of the quantitative data, though, came from two especially thorough studies done for the Zion and Indian Point power plants, operated by Commonwealth Edison and Consolidated Edison respectively. The principal conclusions of the Garrick survey are that the public risk (from radiation release) is now known to be very small for commercial nuclear power plants, but that the risk to utilities (from core damage) is somewhat larger. Significant radiation releases require both core meltdown -- an event occurring only about once every 10,000 reactor-years -- and containment failure, occurring only about once in every hundred meltdowns

MELPROG/TRAC: update and applications

International Nuclear Information System (INIS)

Henninger, R.J.; Kelly, J.E.

1986-01-01

The first complete, coupled, and mechanistic analysis of a core meltdown sequence has been made with MELPROG-PWR/MOD1 and MELPROG/TRAC. The sequence analyzed was a station blackout accident for the Surry plant. Through vessel failure, all important aspects of the meltdown sequence were calculated. This version of MELPROG permits a full two-dimensional treatment of the in-vessel phenomena. Natural circulation can thus be modeled. Comparison to one-dimensional MELPROG and MARCH calculations shows that natural circulation reduces the rate of core heating, but increases the rate of heating of upper plenum structures and primary piping. This increased heating can inhibit fission product deposition and may lead to an early failure of the primary system. Because of uncertainty, sensitivity studies were performed to assess the relative importance of modeling assumptions. Changes in the modeling of the initial fuel rod melting and relocation were found to vary the hydrogen source by a factor of 2 and alter the timing of key events. These results imply that accurate and mechanistic modeling is important for severe accident sequence analysis

The Scenario Model Intercomparison Project (ScenarioMIP) for CMIP6

Energy Technology Data Exchange (ETDEWEB)

O' Neill, Brian C.; Tebaldi, Claudia; van Vuuren, Detlef P.; Eyring, Veronika; Friedlingstein, Pierre; Hurtt, George; Knutti, Reto; Kriegler, Elmar; Lamarque, Jean-Francois; Lowe, Jason; Meehl, Gerald A.; Moss, Richard; Riahi, Keywan; Sanderson, Benjamin M.

2016-01-01

Projections of future climate change play a fundamental role in improving understanding of the climate system as well as characterizing societal risks and response options. The Scenario Model Intercomparison Project (ScenarioMIP) is the primary activity within Phase 6 of the Coupled Model Intercomparison Project (CMIP6) that will provide multi-model climate projections based on alternative scenarios of future emissions and land use changes produced with integrated assessment models. In this paper, we describe ScenarioMIP's objectives, experimental design, and its relation to other activities within CMIP6. The ScenarioMIP design is one component of a larger scenario process that aims to facilitate a wide range of integrated studies across the climate science, integrated assessment modeling, and impacts, adaptation, and vulnerability communities, and will form an important part of the evidence base in the forthcoming Intergovernmental Panel on Climate Change (IPCC) assessments. At the same time, it will provide the basis for investigating a number of targeted science and policy questions that are especially relevant to scenario-based analysis, including the role of specific forcings such as land use and aerosols, the effect of a peak and decline in forcing, the consequences of scenarios that limit warming to below 2 °C, the relative contributions to uncertainty from scenarios, climate models, and internal variability, and long-term climate system outcomes beyond the 21st century. To serve this wide range of scientific communities and address these questions, a design has been identified consisting of eight alternative 21st century scenarios plus one large initial condition ensemble and a set of long-term extensions, divided into two tiers defined by relative priority. Some of these scenarios will also provide a basis for variants planned to be run in other CMIP6-Endorsed MIPs to investigate questions related to specific forcings. Harmonized, spatially

A core framework and scenario for deep GHG reductions at the city scale

International Nuclear Information System (INIS)

Lazarus, Michael; Chandler, Chelsea; Erickson, Peter

2013-01-01

Trends in increasing urbanization, paired with a lack of ambitious action on larger scales, uniquely position cities to resume leadership roles in climate mitigation. While many cities have adopted ambitious long-term emission reduction goals, few have articulated how to reach them. This paper presents one of the first long-term scenarios of deep greenhouse gas abatement for a major U.S. city. Using a detailed, bottom-up scenario analysis, we investigate how Seattle might achieve its recently stated goal of carbon neutrality by the year 2050. The analysis demonstrates that a series of ambitious strategies could achieve per capita GHG reductions of 34% in 2020, and 91% in 2050 in Seattle's “core” emissions from the buildings, transportation, and waste sectors. We examine the pros and cons of options to get to, or beyond, net zero emissions in these sectors. We also discuss methodological innovations for community-scale emissions accounting frameworks, including a “core” emissions focus that excludes industrial activity and a consumption perspective that expands the emissions footprint and scope of policy solutions. As in Seattle, other communities may find the mitigation strategies and analytical approaches presented here are useful for crafting policies to achieve deep GHG-reduction goals. - Highlights: ► Cities can play a pivotal role in mitigating climate change. ► Strategies modeled achieve per-capita GHG reductions of 91% by 2050 in Seattle. ► We discuss methodological innovations in community-scale accounting frameworks. ► We weigh options for getting to, or beyond, zero GHG emissions. ► Other cities may adapt these measures and analytical approaches to curb emissions

Making use of scenarios : supporting scenario use in product design

NARCIS (Netherlands)

Anggreeni, Irene

2010-01-01

The discipline of Scenario-Based Product Design (SBPD) guides the use of scenarios in a product design process. As concrete narratives, scenarios could facilitate making explicit how users would use the designed product in their activities, allowing usability studies to be an integrated part of the

Development of steady-state scenarios compatible with ITER-like wall conditions

Energy Technology Data Exchange (ETDEWEB)

Litaudon, X [Association Euratom-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Arnoux, G [Association Euratom-CEA, CEA/DSM/DRFC-Cadarache 13108, St Paul Durance (France); Beurskens, M [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)] (and others)

2007-12-15

A key issue for steady-state tokamak operation is to determine the edge conditions that are compatible both with good core confinement and with the power handling and plasma exhaust capabilities of the plasma facing components (PFCs) and divertor systems. A quantitative response to this open question will provide a robust scientific basis for reliable extrapolation of present regimes to an ITER compatible steady-state scenario. In this context, the JET programme addressing steady-state operation is focused on the development of non-inductive, high confinement plasmas with the constraints imposed by the PFCs. A new beryllium main chamber wall and tungsten divertor together with an upgrade of the heating/fuelling capability are currently in preparation at JET. Operation at higher power with this ITER-like wall will impose new constraints on non-inductive scenarios. Recent experiments have focused on the preparation for this new phase of JET operation. In this paper, progress in the development of advanced tokamak (AT) scenarios at JET is reviewed keeping this long-term objective in mind. The approach has consisted of addressing various critical issues separately during the 2006-2007 campaigns with a view to full scenario integration when the JET upgrades are complete. Regimes with internal transport barriers (ITBs) have been developed at q{sub 95} {approx} 5 and high triangularity, {delta} (relevant to the ITER steady-state demonstration) by applying more than 30 MW of additional heating power reaching {beta}{sub N} {approx} 2 at B{sub o} {approx} 3.1 T. Operating at higher {delta} has allowed the edge pedestal and core densities to be increased pushing the ion temperature closer to that of the electrons. Although not yet fully integrated into a performance enhancing ITB scenario, Neon seeding has been successfully explored to increase the radiated power fraction (up to 60%), providing significant reduction of target tile power fluxes (and hence temperatures) and

Results of research and development works 1982

International Nuclear Information System (INIS)

1983-02-01

Arranged by main topics of research, the following contributions of the Institute are disposed as follows: 1. Fast Breeder Reactor Project (PSB) 1.1 Nuclear fuels and fuel elements 1.2 Cladding and core structure materials 1.3 Plant structure materials 1.4 Safety investigations 2. Nuclear Safety Project (PNS) 2.1 Fuel element behaviour during accidents 2.2 Core meltdown 3. Reprocessing and Waste Management Project (PWA) 4. Fusion technology (FT) 5. Innovations and new tasks (INNA). (orig./IHOe) [de

Dying scenarios improve recall as much as survival scenarios.

Science.gov (United States)

Burns, Daniel J; Hart, Joshua; Kramer, Melanie E

2014-01-01

Merely contemplating one's death improves retention for entirely unrelated material learned subsequently. This "dying to remember" effect seems conceptually related to the survival processing effect, whereby processing items for their relevance to being stranded in the grasslands leads to recall superior to that of other deep processing control conditions. The present experiments directly compared survival processing scenarios with "death processing" scenarios. Results showed that when the survival and dying scenarios are closely matched on key dimensions, and possible congruency effects are controlled, the dying and survival scenarios produced equivalently high recall levels. We conclude that the available evidence (cf. Bell, Roer, & Buchner, 2013; Klein, 2012), while not definitive, is consistent with the possibility of overlapping mechanisms.

POLCA-T simulation of OECD/NRC BWR turbine trip benchmark exercise 3 best estimate scenario TT2 test and four extreme scenarios

International Nuclear Information System (INIS)

Panayotov, D.

2004-01-01

Westinghouse transient code POLCA-T brings together the system thermal-hydraulics plant models and the 3D neutron kinetics core model. Code validation plan includes the calculations of Peach Bottom end of cycle 2 turbine trip transients and low-flow stability tests. The paper describes the objectives, method, and results of analyses performed in the final phase of OECD/NRC Peach Bottom 2 Boiling Water Reactor Turbine Trip Benchmark. Brief overview of the code features, the method of simulation, the developed 3D core model and system input deck for Peach Bottom 2 are given. The paper presents the results of benchmark exercise 3 best estimate scenario: coupled 3D core neutron kinetics with system thermal-hydraulics analyses. Performed sensitivity studies cover the SCRAM initiation, carry-under, and decay power. Obtained results including total power, steam dome, core exit, lower and upper plenum, main steam line and turbine inlet pressures showed good agreement with measured plant data Thus the POLCA-T code capabilities for correct simulation of turbine trip transients were proved The performed calculations and obtained results for extreme cases demonstrate the POLCA-T code wide range capabilities to simulate transients when scram, steam bypass, and safety and relief valves are not activated. The code is able to handle such transients even when the reactor power and pressure reach values higher than 600 % of rated power, and 10.8 MPa. (authors)

Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

International Nuclear Information System (INIS)

Perez, S.S.

1987-01-01

The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

A status report on the integral fast reactor fuels and safety program

International Nuclear Information System (INIS)

Pedersen, D.R.; Seidel, B.R.

1990-01-01

The integral fast reactor (IFR) is an advanced liquid-metal-cooled reactor (ALMR) concept being developed at Argonne National Laboratory. The IFR program is specifically responsible for the irradiation performance, advanced core design, safety analysis, and development of the fuel cycle for the US Department of Energy's ALMR program. The basic elements of the IFR concept are (a) metallic fuel, (b) liquid-sodium cooling, (c) modular, pool-type reactor configuration, (d) an integral fuel cycle based upon pyrometallurgical processing. The most significant safety aspects of the IFR program result from its unique fuel design, a ternary alloy of uranium, plutonium, and zirconium. This fuel is based on experience gained through > 25 yr operation of the Experimental Breeder Reactor II (EBR-II) with a uranium alloy metallic fuel. The ultimate criteria for fuel pin design is the overall integrity at the target burnup. The probability of core meltdown is remote; however, a theoretical possibility of core meltdown remains. The next major step in the IFR development program will be a full-scale pyroprocessing demonstration to be carried out in conjunction with EBR-II. The IFR fuel cycle closure based on pyroprocessing will also have a dramatic impact on waste management options and on actinide recycling

Analysis of the forever-experiments using a finite element model

International Nuclear Information System (INIS)

Altstadt, E.; Moessner, T.; Sehgal, B.R.; Willschuetz, H.G.

2000-01-01

For future nuclear power plants it is demanded that there are no consequences for the environment and the population even in the closest vicinity of the plant during and after every possible accident scenario. This includes 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). One accident management strategy is to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. This strategy also applies to existing plants. Therefore, an axis-symmetric Finite Element (FE) model is developed based on the multipurpose code ANSYS/Multiphysics trademark . Using the Computational Fluid Dynamics (CFD) module the temperature field within the melt pool and within the vessel wall is evaluated. The transient structural mechanical calculations are then performed applying a creep model which takes into account large temperature, stress and strain variations. Taking into account both - experimental and numerical results - provides a good opportunity to improve the simulation and the understanding of the vessel failure mechanisms. Of particular interest are (i) the time to failure and (ii) the location and mode of failure. (orig.)

EDITORIAL: Where next with global environmental scenarios? Where next with global environmental scenarios?

Science.gov (United States)

O'Neill, Brian; Pulver, Simone; Van Deveer, Stacy; Garb, Yaakov

2008-12-01

Scenarios have become a standard tool in the portfolio of techniques that scientists and policy-makers use to envision and plan for the future. Defined as plausible, challenging and relevant stories about how the future might unfold that integrate quantitative models with qualitative assessments of social and political trends, scenarios are a central component in assessment processes for a range of global issues, including climate change, biodiversity, agriculture, and energy. Yet, despite their prevalence, systematic analysis of scenarios is in its beginning stages. Fundamental questions remain about both the epistemology and scientific credibility of scenarios and their roles in policymaking and social change. Answers to these questions have the potential to determine the future of scenario analyses. Is scenario analysis moving in the direction of earth system governance informed by global scenarios generated through increasingly complex and comprehensive models integrating socio-economic and earth systems? Or will global environmental scenario analyses lose favour compared to more focused, policy-driven, regionally specific modelling? These questions come at an important time for the climate change issue, given that the scenario community, catalyzed by the Intergovernmental Panel on Climate Change (IPCC), is currently preparing to embark on a new round of scenario development processes aimed at coordinating research and assessment, and informing policy, over the next five to ten years. These and related questions about where next to go with global environmental scenarios animated a workshop held at Brown University (Note1) that brought together leading practitioners and scholars of global environmental change scenarios from research, policy-making, advocacy, and business settings. The workshop aimed to provide an overview of current practices/best practices in scenario production and scenario use across a range of global environmental change arenas. Participants

ATHENA-2D: A computer code for simulation of hypothetical recriticality accidents in a thermal neutron spectrum

International Nuclear Information System (INIS)

Schwinkendorf, K.N.

1995-01-01

In a damaged light water reactor core (as in the aftermath of a Three-Mile-Island-like core meltdown), water reflood is performed to carry off decay heat. The severely degraded geometry of the fuel debris bed may increase core reactivity with water reflood. Sufficient boron poisoning of the reflood water is therefore very important. One hypothetical accident is the reintroduction of cooling water that is insufficiently borated, resulting in the damaged reactor attaining criticality in this uncontrolled configuration. The goal in simulating this accident is the prediction of the energy release from the resulting transient

Induced supersolidity in a mixture of normal and hard-core bosons

International Nuclear Information System (INIS)

Mishra, Tapan; Das, B. P.; Pai, Ramesh V.

2010-01-01

We present a scenario where a supersolid is induced in one of the components of a mixture of two species bosonic atoms where there are no long-range interactions. We study a system of normal and hard-core boson mixture with only the former possessing long-range interactions. We consider three cases: the first where the total density is commensurate and the other two where it is incommensurate to the lattice. By suitable choices of the densities of normal and hard-core bosons and the interaction strengths between them, we predict that the charge density wave and the supersolid orders can be induced in the hard-core species as a result of the competing interatomic interactions.

Study of nuclear energy systems and double strata scenarios for minor actinides transmutation in ADS

International Nuclear Information System (INIS)

Clavel, J.B.

2012-01-01

The French law of 28 June 2006 regarding advanced nuclear waste management requires a scientific assessment to define future industrial strategies. The present PhD thesis was carried in this framework and concerns specifically the research axis of minor actinides transmutation. A high power Accelerator Driven System (ADS) concept is developed at SUBATECH for this purpose. A 1 GeV proton beam feeds three liquid lead-bismuth spallation targets. The Multiple Spallation Target (MUST) ADS reaches the thermal powers up to 1 GW with a high specific power. A nuclear reactor dimensioning method has been developed and applied to different double strata scenarios. In these scenarios, SFR (Sodium Fast Reactors) or PWR (Pressurized Water Reactors) power reactors produce minor actinides that will be transmuted into ADS. In each core (SFR and ADS), the plutonium multi-reprocessing strategy is performed while ADS subcritical core also multi-reprocesses minor actinides. To limit the core reactivity and improve the fuel thermal conductivity, the minor actinides fuel is mixed with MgO inert matrix. Nuclear branches with lead and sodium coolants for the ADS, have been studied for different irradiation times and two transmutation strategies have been assessed: whether whole minor actinides, whether americium only is transmuted. The thesis presents precisely the MUST ADS design methodology and the calculations to get a fuel composition at equilibrium. Then a one cycle evolution is performed and analysed for the fuel and the multiplication factor. Radiotoxicity and thermal power of the waste produced are then compared. Finally, the study of double strata scenarios is performed to analyse the plutonium and minor actinides inventories in cycle and also the waste produced according to the transmutation strategies applied and the first stratum evolution. (author)

The world in scenarios

International Nuclear Information System (INIS)

De Jong, A.; Roodenburg, H.

1992-01-01

As an introduction to this special issue 'Worlds of difference: Scenarios's for the economy, energy and the environment 1990-2015', an outline is given of the future of the world and the Netherlands, based on four scenarios. These scenarios are published in 'Scanning the future' in May 1992 by the CPB, the Dutch Central Planning Bureau. The Global Shift (GS) scenario is characterized by a very dynamic technological development, the free market perspective, strong economic growth in the Asian economies, and a relative economic regression in Western Europe. In the European Renaissance (ER) scenario the technological development is less dynamic and more gradual than in the GS scenario. The Balanced Growth (BG) scenario is dominated by a sustainable economic development and a strong technological dynamic development. The Global Crisis (GC) scenario shows a downward spiral in many areas, stagnating developments and fragile economies as results of the trends in the eighties. The first three scenarios are elaborated for the Netherlands. Also attention is paid to the aims and meaning of long-term scenarios. 2 figs., 2 tabs., 3 refs

An Estimation of Operator's Diagnostic Time for Feed-And-Bleed Operation under Various Scenarios

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Abu Dhabi (United Arab Emirates)

2015-10-15

In order to assess realistic safety of plant, effects of interactions between components, operator, and plant condition are needed to be considered in the PSA model. One of the important issues to estimate the CDF is the estimation of human error probability (HEP). When an accident occurs, operators follow the emergency operating procedure and check various alarm, parameters, and signals. In the conventional Korean PSA model, the Korean standard HRA (K-HRA) method is used. In this method, the HEP is the sum of diagnosis error probability and execution error probability. A diagnosis error probability is expressed by the available time for diagnosis and adjusting performance shaping factors, and an execution error probability is a function of task type and stress level. Available time for diagnosis is very important factor of HEP. If the available time for diagnosis is short, the HEP becomes high. In order to obtain the realistic risk assessment results, we first focus on the estimation of HEP considering the plant dynamics under various scenarios. Target operation and scenarios are feed-and bleed operation (F and B operation) and total loss of feedwater (TLOFW) accident with/without loss of coolant accident (LOCA). One of the highest HEP is HEP of FnB operation. In additional, Scenarios, which are related to combination secondary heat removal failure and primary heat removal failure, are most critical core damage scenario of the combined accident except scenarios related to station black out (SBO). In these scenarios, the FnB operation is last resort to prevent core damage. To estimate available operator diagnosis time, we identify the relationship between accidents, mitigation function, and plant condition. Distribution of available time of diagnosis was estimated using the MOSAIQUE. The variables are break size, break timing, trip timing of RCP, and availability of high pressure safety injection (HPSI) pump under the TLOFW accident with LOCA. For Type 1 accident

The simultaneous measurements of core and outer core density fluctuations in L-H transition using CO2 laser collective scattering diagnostic in the EAST superconducting tokamak

International Nuclear Information System (INIS)

Cao, G.M.; Li, Y.D.; Zhang, X.D.; Sun, P.J.; Hu, L.Q.; Li, J.G.; Wu, G.J.

2013-01-01

The H-mode is the projected basic operation scenario for the ITER tokamak. The turbulence de-correlation by the synergistic effect of zonal flow and equilibrium ExB flow shear is believed to be the reason for L-H transition, however, the detailed physical mechanism has not been identified so far. Tangential multi-channel CO 2 laser collective scattering diagnostic system (mainly k r measurement) was first installed to investigate electron density fluctuations on EAST tokamak. The measurements in a spontaneous dithering L-H transition show that in core plasma (0 < r/a < 0.5) the low-frequency fluctuations strengthen greatly before L-H transition; meanwhile in outer core plasma (0.2 < r/a < 1) the low-frequency fluctuations strengthen slightly. Bispectral analysis reveals that the coupling strength between low- and high-frequency fluctuations in both core and outer core plasma strengthens greatly before the transition, but the latter is greater than the former. The results indicate that the low-frequency fluctuations of the core and outer core plasma play active, but different, roles in the spontaneous L-H transition. (author)

Verification of CTF/PARCSv3.2 coupled code in a Turbine Trip scenario

International Nuclear Information System (INIS)

Abarca, A.; Hidalga, P.; Miro, R.; Verdu, G.; Sekhri, A.

2017-01-01

Multiphysics codes had revealed as a best-estimate approach to simulate core behavior in LWR. Coupled neutronics and thermal-hydraulics codes are being used and improved to achieve reliable results for reactor safety transient analysis. The implementation of the feedback procedure between the coupled codes at each time step allows a more accurate simulation and a better prediction of the safety limits of analyzed scenarios. With the objective of testing the recently developed CTF/PARCSv3.2 coupled code, a code-to-code verification against TRACE has been developed in a BWR Turbine Trip scenario. CTF is a thermal-hydraulic subchannel code that features two-fluid, three-field representation of the two-phase flow, while PARCS code solves the neutronic diffusion equation in a 3D nodal distribution. PARCS features allow as well the use of extended sets of cross section libraries for a more precise neutronic performance in different formats like PMAX or NEMTAB. Using this option the neutronic core composition of KKL will be made taking advantage of the core follow database. The results of the simulation will be verified against TRACE results. TRACE will be used as a reference code for the validation process since it has been a recommended code by the USNRC. The model used for TRACE includes a full core plus relevant components such as the steam lines and the valves affecting and controlling the turbine trip evolution. The coupled code performance has been evaluated using the Turbine Trip event that took place in Kern Kraftwerk Leibstadt (KKL), at the fuel cycle 18. KKL is a Nuclear Power Plant (NPP) located in Leibstadt, Switzerland. This NPP operates with a BWR developing 3600 MWt in fuel cycles of one year. The Turbine Trip is a fast transient developing a pressure peak in the reactor followed by a power decreasing due to the selected control rod insertion. This kind of transient is very useful to check the feedback performance between both coupled codes due to the fast

Overview description of the base scenario derived from FEP analysis

International Nuclear Information System (INIS)

Locke, J.; Bailey, L.

1998-01-01

This report forms part of a suite of documents describing the Nirex model development programme. The programme is designed to provide a clear audit trail from the identification of significant features, events and processes (FEPs) to the models and modelling processes employed within a detailed performance assessment. A scenario approach to performance assessment has been adopted. It is proposed that potential evolutions of a deep geological radioactive waste repository can be represented by a base scenario and a number of variant scenarios. It is intended that assessment of the base scenario would form the core of any future performance assessment. The base scenario is chosen to be broad-ranging and to represent the natural evolution of the repository system and its surrounding environment. The base scenario is defined to include all those FEPs which are certain to occur and those which are judged likely to occur for a significant period of the assessment - timescale. Variant scenarios are defined by FEPs which represent a significant perturbation to the natural system evolution, for example the occurrence of a large seismic event. The structuring of FEPs on a Master Directed Diagram (MDD), provides a systematic framework for identifying those FEPs which form part of the natural evolution of the system and those which may define alternative potential evolutions of the repository system. In order to construct a description of the base scenario, FEPs have been grouped into a series of conceptual models. Conceptual models are groups of FEPs, identified from the MDD, representing a specific component or process within the disposal system. It has been found appropriate to define conceptual models in terms of the three main components of the disposal system: the repository engineered system, the surrounding geosphere and the biosphere. For each of these components, conceptual models provide a description of the relevant subsystem in terms of its initial characteristics

Carbon Solubility in Silicon-Iron-Bearing Metals during Core Formation on Mercury

Science.gov (United States)

Vander Kaaden, Kathleen E.; McCubbin, Francis M.; Ross, D. Kent; Rapp, Jennifer F.; Danielson, Lisa R.; Keller, Lindsay P.; Righter, Kevin

2016-01-01

Recent results obtained from the MErcury Surface, Space ENvironment, GEochemistry, and Ranging (MESSENGER) spacecraft showed the surface of Mercury has high S abundances (approximately 4 wt%) and low Iron(II) Oxide abundances (less than 2 wt%). Based on these extreme values, the oxygen fugacity of Mercury's surface materials was estimated to be approximately 3 to 7 log(sub 10) units below the IW buffer (Delta IW-3 to Delta IW-7). This highly reducing nature of the planet has resulted in a large core and relatively thin mantle, extending to only approximately 420 km depth (corresponding to a core-mantle boundary pressure of approximately 4-7 GPa) within the planet. Furthermore, MESSENGER results have suggested the presence of carbon on the surface of the planet. Previous experimental results from have also suggested the possibility of a primary floatation crust on Mercury composed of graphite, produced after a global magma ocean event. With these exotic conditions of this compositional end-member planet, it begs the question, what is the core composition of Mercury? Although no definitive conclusion has been reached, previous studies have made advances towards answering this question. Riner et al. and Chen et al. looked at iron sulfide systems and implemented various crystallization and layered core scenarios to try and determine the composition and structure of Mercury's core. Malavergne et al. examined core crystallization scenarios in the presence of sulfur and silicon. Hauck et al. used the most recent geophysical constraints from the MESSENGER spacecraft to model the internal structure of Mercury, including the core, in a iron-sulfur-silicon system. More recently, Chabot et al. conducted a series of metal-silicate partitioning experiments in a iron-sulfur-silicon system. These results showed the core of Mercury has the potential to contain more than 15 wt% silicon. However, with the newest results from MESSENGER's low altitude campaign, carbon is another

Ion cyclotron resonance heating for tungsten control in various JET H-mode scenarios

Science.gov (United States)

Goniche, M.; Dumont, R. J.; Bobkov, V.; Buratti, P.; Brezinsek, S.; Challis, C.; Colas, L.; Czarnecka, A.; Drewelow, P.; Fedorczak, N.; Garcia, J.; Giroud, C.; Graham, M.; Graves, J. P.; Hobirk, J.; Jacquet, P.; Lerche, E.; Mantica, P.; Monakhov, I.; Monier-Garbet, P.; Nave, M. F. F.; Noble, C.; Nunes, I.; Pütterich, T.; Rimini, F.; Sertoli, M.; Valisa, M.; Van Eester, D.; Contributors, JET

2017-05-01

Ion cyclotron resonance heating (ICRH) in the hydrogen minority scheme provides central ion heating and acts favorably on the core tungsten transport. Full wave modeling shows that, at medium power level (4 MW), after collisional redistribution, the ratio of power transferred to the ions and the electrons vary little with the minority (hydrogen) concentration n H/n e but the high-Z impurity screening provided by the fast ions temperature increases with the concentration. The power radiated by tungsten in the core of the JET discharges has been analyzed on a large database covering the 2013-2014 campaign. In the baseline scenario with moderate plasma current (I p = 2.5 MA) ICRH modifies efficiently tungsten transport to avoid its accumulation in the plasma centre and, when the ICRH power is increased, the tungsten radiation peaking evolves as predicted by the neo-classical theory. At higher current (3-4 MA), tungsten accumulation can be only avoided with 5 MW of ICRH power with high gas injection rate. For discharges in the hybrid scenario, the strong initial peaking of the density leads to strong tungsten accumulation. When this initial density peaking is slightly reduced, with an ICRH power in excess of 4 MW,very low tungsten concentration in the core (˜10-5) is maintained for 3 s. MHD activity plays a key role in tungsten transport and modulation of the tungsten radiation during a sawtooth cycle is correlated to the fishbone activity triggered by the fast ion pressure gradient.

Food scenarios 2025

DEFF Research Database (Denmark)

Sundbo, Jon

2016-01-01

This article presents the results of a future study of the food sector. Two scenarios have been developed using a combination of: 1) a summary of the relevant scientific knowledge, 2) systematic scenario writing, 3) an expert-based Delphi technique, and 4) an expert seminar assessment. The two...... scenarios present possible futures at global, national (Denmark) and regional (Zealand, Denmark) levels. The main scenario is called ‘Food for ordinary days and celebrations’ (a combination of ‘High-technological food production − The functional society’ and ‘High-gastronomic food − The experience society...

Interpreting energy scenarios

Science.gov (United States)

Iyer, Gokul; Edmonds, James

2018-05-01

Quantitative scenarios from energy-economic models inform decision-making about uncertain futures. Now, research shows the different ways these scenarios are subsequently used by users not involved in their initial development. In the absence of clear guidance from modellers, users may place too much or too little confidence in scenario assumptions and results.

Global warming: Climate scenarios and international agriculture

International Nuclear Information System (INIS)

Downing, T.E.; Parry, M.L.

1991-01-01

The potential impacts of climatic change on international agriculture are summarized, drawing on results from the Intergovernmental Panel on Climate Change impacts working group. The four different climate change scenarios used for investigating impacts: historical studies, artificial scenarios, analogues, and general circulation models, are briefly reviewed. Climate change will affect agriculture in three ways: direct effects of increased carbon dioxide concentration, effects of altered weather patterns, and secondary effects on social and economic situations. The effect of increased carbon dioxide concentration is uncertain, but potentially will enhance plant growth and water use efficiency. The sensitivity of grain maize to incremental changes in annual temperature is described, with the suitable zone expanding from the middle of Europe to southern Scandinavia. Potential damage from insect pests may increase under warmer climates, with northerly movement of insect breeding grounds. Temperature increases are likely to lengthen the growing season where temperature is a limiting factor, especially at higher lattitudes in the Northern Hemisphere. Higher temperatures, shorter periods of grain filling, and reduced winter chilling will reduce potential yields in current core grain-growing areas, and changing moisture regimes will shift agricultural patterns. The horn of Africa and parts of western Africa are likely to suffer enhanced food supply vulnerability. 16 refs., 4 figs

Hydromagnetic instabilities and magnetic field amplification in core collapse supernovae

Energy Technology Data Exchange (ETDEWEB)

Cerda-Duran, P; Obergaulinger, M; Mueller, E [Max-Planck-Institut fuer Astrophysik, Karl-Schwarzschild-st. 1, 85748 Garching (Germany); Aloy, M A; Font, J A, E-mail: cerda@mpa-garching.mpg.de [Departamento de Astronomia y Astrofisica, Universidad de Valencia, 46100 Burjassot, Valencia (Spain)

2011-09-22

Some of the most violent events in the universe, the gamma ray burst, could be related to the gravitational collapse of massive stellar cores. The recent association of long GRBs to some class of type Ic supernova seems to support this view. In such scenario fast rotation, strong magnetic fields and general relativistic effects are key ingredients. It is thus important to understand the mechanism that amplifies the magnetic field under that conditions. I present global simulations of the magneto-rotational collapse of stellar cores in general relativity and semi-global simulations of hydromagnetic instabilities under core collapse conditions. I discuss effect of the magneto-rotational instability and the magnetic field amplification during the collapse, the uncertainties in this process and the dynamical effects in the supernova explosion.

HD271791: dynamical versus binary-supernova ejection scenario

Science.gov (United States)

Gvaramadze, V. V.

2009-05-01

The atmosphere of the extremely high-velocity (530-920kms-1) early B-type star HD271791 is enriched in α-process elements, which suggests that this star is a former secondary component of a massive tight binary system and that its surface was polluted by the nucleosynthetic products after the primary star exploded in a supernova. It was proposed that the (asymmetric) supernova explosion unbind the system and that the secondary star (HD271791) was released at its orbital velocity in the direction of Galactic rotation. In this Letter, we show that to explain the Galactic rest-frame velocity of HD271791 within the framework of the binary-supernova scenario, the stellar remnant of the supernova explosion (a =750-1200kms-1. We therefore consider the binary-supernova scenario as highly unlikely and instead propose that HD271791 attained its peculiar velocity in the course of a strong dynamical three- or four-body encounter in the dense core of the parent star cluster. Our proposal implies that by the moment of encounter HD271791 was a member of a massive post-supernova binary.

Analysis of the return to power scenario following a LBLOCA in a PWR

Energy Technology Data Exchange (ETDEWEB)

Macian, R.; Tyler, T.N.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

1995-09-01

The risk of reactivity accidents has been considered an important safety issue since the beginning of the nuclear power industry. In particular, several events leading to such scenarios for PWR`s have been recognized and studied to assess the potential risk of fuel damage. The present paper analyzes one such event: the possible return to power during the reflooding phase following a LBLOCA. TRAC-PF1/MOD2 coupled with a three-dimensional neutronic model of the core based on the Nodal Expansion Method (NEM) was used to perform the analysis. The system computer model contains a detailed representation of a complete typical 4-loop PWR. Thus, the simulation can follow complex system interactions during reflooding, which may influence the neutronics feedback in the core. Analyses were made with core models bases on cross sections generated by LEOPARD. A standard and a potentially more limiting case, with increased pressurizer and accumulator inventories, were run. In both simulations, the reactor reaches a stable state after the reflooding is completed. The lower core region, filled with cold water, generates enough power to boil part of the incoming liquid, thus preventing the core average liquid fraction from reaching a value high enough to cause a return to power. At the same time, the mass flow rate through the core is adequate to maintain the rod temperature well below the fuel damage limit.

Development and application of a methodology for identifying and characterising scenarios

International Nuclear Information System (INIS)

Billington, D.; Bailey, L.

1998-01-01

This report forms part of a suite of documents describing the Nirex model development programme. The programme is designed to provide a clear audit trail from the identification of significant features, events and processes (FEPs) to the models and modelling processes employed within a detailed performance assessment. A scenario approach to performance assessment has been adopted. It is proposed that potential evolutions of a deep geological radioactive waste repository can be represented by a base scenario and a number of variant scenarios. It is intended that assessment of the base, scenario would form the core of any future performance assessment. The base scenario is chosen to be broad-ranging and to represent the natural evolution of the repository system and its surrounding environment. The base scenario is defined to include all those FEPs which are certain to occur and those which are judged likely to occur for a significant period of the assessment timescale. Variant scenarios are defined by FEPs which represent a significant perturbation to the natural system evolution, for example the occurrence of a large seismic event. A variant scenario defined by a single initiating FEP is characterised by a sequence of events. This is represented as a 'timeline' which forms the basis for modelling that scenario. To generate a variant scenario defined by two initiating FEPs, a methodology is presented for combining the timelines for the two underlying 'single-FEP' variants. The resulting series of event sequences can be generated automatically. These sequences are then reviewed, in order to reduce the number of timelines requiring detailed consideration. This is achieved in two ways: by aggregating sequences which have similar consequence in terms of safety performance; and by combining successive intervals along a timeline where appropriate. In the context of a performance assessment, the aim is to determine the conditional risk and appropriate weight for each

The role of the core in degeneracy of chiral candidate band doubling

International Nuclear Information System (INIS)

Timar, J.; Sohler, D.; Vaman, C.; SUNY, Stony Brook, NY; Starosta, K.; Fossan, D.B.; Koike, T.; Tohoku Univ., Sendai; Lee, I.Y.; Macchiavelli, A.O.

2005-01-01

Complete text of publication follows. Nearly degenerate ΔI=1 rotational bands have been observed recently in several odd-odd nuclei in the A ∼ 130 and A ∼ 100 mass regions. The properties of these doublet bands have been found to agree with the scenario of spontaneous formation of chirality and disagree with other possible scenarios. However, the most recent results obtained from life-time experiments for some chiral candidate nuclei in the A ∼ 130 mass region seem to contradict the chiral interpretation of the doublet bands in these nuclei based on the observed differences in the absolute electromagnetic transition rates; the transition rates expected for chiral doublets are predicted to be very similar. Therefore it is interesting to search for new types of experimental data that may provide further possibilities to distinguish between alternative interpretations, and may uncover new properties of the mechanism that is responsible for the band doubling in these nuclei. Such a new type of experimental data was found by studying the chiral candidate bands in neighboring Rh nuclei. High-spin states of 103 Rh were studied using the 96 Zr( 11 B,4n) reaction at 40 MeV beam energy and chiral partner candidate bands have been found in it. As a result of this observation a special quartet of neighboring chiral candidate nuclei can be investigated for the first time. With this quartet identified a comparison between the behavior of the nearly degenerate doublet bands belonging to the same core but to different valence quasiparticle configurations, as well as belonging to different cores but to the same valence quasiparticle configuration, becomes possible. The comparison shows that the energy separation of these doublet band structures depends mainly on the core properties and only at less extent on the valence quasiparticle coupling. This observation sets up new criteria for the explanations of the band doublings, restricting the possible scenarios and providing

MAKE SUPER-EARTHS, NOT JUPITERS: ACCRETING NEBULAR GAS ONTO SOLID CORES AT 0.1 AU AND BEYOND

Energy Technology Data Exchange (ETDEWEB)

Lee, Eve J.; Chiang, Eugene; Ormel, Chris W., E-mail: evelee@berkeley.edu, E-mail: echiang@astro.berkeley.edu, E-mail: ormel@berkeley.edu [Department of Astronomy, University of California Berkeley, Berkeley, CA 94720-3411 (United States)

2014-12-20

Close-in super-Earths having radii 1-4 R {sub ⊕} may possess hydrogen atmospheres comprising a few percent by mass of their rocky cores. We determine the conditions under which such atmospheres can be accreted by cores from their parent circumstellar disks. Accretion from the nebula is problematic because it is too efficient: we find that 10 M {sub ⊕} cores embedded in solar metallicity disks tend to undergo runaway gas accretion and explode into Jupiters, irrespective of orbital location. The threat of runaway is especially dire at ∼0.1 AU, where solids may coagulate on timescales orders of magnitude shorter than gas clearing times; thus nascent atmospheres on close-in orbits are unlikely to be supported against collapse by planetesimal accretion. The time to runaway accretion is well approximated by the cooling time of the atmosphere's innermost convective zone, whose extent is controlled by where H{sub 2} dissociates. Insofar as the temperatures characterizing H{sub 2} dissociation are universal, timescales for core instability tend not to vary with orbital distance—and to be alarmingly short for 10 M {sub ⊕} cores. Nevertheless, in the thicket of parameter space, we identify two scenarios, not mutually exclusive, that can reproduce the preponderance of percent-by-mass atmospheres for super-Earths at ∼0.1 AU, while still ensuring the formation of Jupiters at ≳ 1 AU. Scenario (a): planets form in disks with dust-to-gas ratios that range from ∼20× solar at 0.1 AU to ∼2× solar at 5 AU. Scenario (b): the final assembly of super-Earth cores from mergers of proto-cores—a process that completes quickly at ∼0.1 AU once begun—is delayed by gas dynamical friction until just before disk gas dissipates completely. Both scenarios predict that the occurrence rate for super-Earths versus orbital distance, and the corresponding rate for Jupiters, should trend in opposite directions, as the former population is transformed into the latter: as

Designing Sustainable Production and Distribution Scenarios for the Beef and Pork Supply Chains in Brazil

Directory of Open Access Journals (Sweden)

Giulia Dondè

2016-08-01

The core improvements that can be extracted from the scenarios are: increased chain collaboration, information sharing in the chain, and governmental involvement in the Brazilian beef and pork supply chains. A major value of the paper lays in the three-phase methodology set-up. This paper will also contribute to the debate on sustainable improvements feasible in these Brazilian food chains.

Integrative Scenario Development

Directory of Open Access Journals (Sweden)

Joerg A. Priess

2014-03-01

Full Text Available Scenarios are employed to address a large number of future environmental and socioeconomic challenges. We present a conceptual framework for the development of scenarios to integrate the objectives of different stakeholder groups. Based on the framework, land-use scenarios were developed to provide a common base for further research. At the same time, these scenarios assisted regional stakeholders to bring forward their concerns and arrive at a shared understanding of challenges between scientific and regional stakeholders, which allowed them to eventually support regional decision making. The focus on the integration of views and knowledge domains of different stakeholder groups, such as scientists and practitioners, required rigorous and repeated measures of quality control. The application of the integrative concept provided products for both stakeholder groups, and the process of scenario development facilitated cooperation and learning within both the scientist and practitioner groups as well as between the two groups.

The future of scenarios: issues in developing new climate change scenarios

International Nuclear Information System (INIS)

Pitcher, Hugh M

2009-01-01

In September, 2007, the IPCC convened a workshop to discuss how a new set of scenarios to support climate model runs, mitigation analyses, and impact, adaptation and vulnerability research might be developed. The first phase of the suggested new approach is now approaching completion. This article discusses some of the issues raised by scenario relevant research and analysis since the last set of IPCC scenarios were created (IPCC SRES, 2000) that will need to be addressed as new scenarios are developed by the research community during the second phase. These include (1) providing a logic for how societies manage to transition from historical paths to the various future development paths foreseen in the scenarios, (2) long-term economic growth issues, (3) the appropriate GDP metric to use (purchasing power parity or market exchange rates), (4) ongoing issues with moving from the broad geographic and time scales of the emission scenarios to the finer scales needed for impacts, adaptation and vulnerability analyses and (5) some possible ways to handle the urgent request from the policy community for some guidance on scenario likelihoods. The challenges involved in addressing these issues are manifold; the reward is greater credibility and deeper understanding of an analytic tool that does much to form the context within which many issues in addition to the climate problem will need to be addressed.

The future of scenarios: issues in developing new climate change scenarios

Science.gov (United States)

Pitcher, Hugh M.

2009-04-01

In September, 2007, the IPCC convened a workshop to discuss how a new set of scenarios to support climate model runs, mitigation analyses, and impact, adaptation and vulnerability research might be developed. The first phase of the suggested new approach is now approaching completion. This article discusses some of the issues raised by scenario relevant research and analysis since the last set of IPCC scenarios were created (IPCC SRES, 2000) that will need to be addressed as new scenarios are developed by the research community during the second phase. These include (1) providing a logic for how societies manage to transition from historical paths to the various future development paths foreseen in the scenarios, (2) long-term economic growth issues, (3) the appropriate GDP metric to use (purchasing power parity or market exchange rates), (4) ongoing issues with moving from the broad geographic and time scales of the emission scenarios to the finer scales needed for impacts, adaptation and vulnerability analyses and (5) some possible ways to handle the urgent request from the policy community for some guidance on scenario likelihoods. The challenges involved in addressing these issues are manifold; the reward is greater credibility and deeper understanding of an analytic tool that does much to form the context within which many issues in addition to the climate problem will need to be addressed.

Physical properties of core-concrete systems: Al{sub 2}O{sub 3}-ZrO{sub 2} molten materials measured by aerodynamic levitation

Energy Technology Data Exchange (ETDEWEB)

Ohishi, Yuji, E-mail: ohishi@see.eng.osaka-u.ac.jp [Graduate School of Engineering, Osaka University (Japan); Kargl, F. [Institute of Materials Physics in Space, German Aerospace Center (Germany); Nakamori, F.; Muta, Hiroaki; Kurosaki, Ken [Graduate School of Engineering, Osaka University (Japan); Yamanaka, Shinsuke [Graduate School of Engineering, Osaka University (Japan); Research Institute of Nuclear Engineering, University of Fukui (Japan)

2017-04-15

During a molten core–concrete interaction, molten oxides consisting of molten core materials (UO{sub 2} and ZrO{sub 2}) and concrete (Al{sub 2}O{sub 3}, SiO{sub 2}, CaO) are formed. Reliable data on the physical properties of the molten oxides will allow us to accurately predict the progression of a nuclear reactor core meltdown accident. In this study, the viscosities and densities of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) were measured using an aerodynamic levitation technique. The densities of two small samples were estimated from their masses and their volumes (calculated from recorded images of the molten samples). The droplets were forced to oscillate using speakers, and their viscosities were evaluated from the damping behaviors of their oscillations. The results showed that the viscosity of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} compared to that of pure molten Al{sub 2}O{sub 3} is 25% lower for x = 0.172, while it is unexpectedly 20% higher for x = 0.356. - Highlights: •The physical properties of molten (ZrO{sub 2}){sub x}(Al{sub 2}O{sub 3}){sub 1-x} (x = 0.356 and 0.172) have been evaluated. •The measurement was conducted using an aerodynamic levitation technique. •The density and viscosity were measured.

A current perspective on the risk significance of steam explosions

International Nuclear Information System (INIS)

Snyder, A.W.

1982-01-01

The view currently held in the Sandia National Laboratory is that, in the case of a meltdown in the reactor core, the probability of a steam explosion is greater than was estimated in WASH-1400, but that the extent and effect of an explosion will be very much smaller than assumed in WASH-1400. This results in a far smaller total risk with regard to containment. In WASH-1400, a nominal conditional probability of 1% was assumed for a containment rupture in a PWR-type reactor, should a large part of the reactor fuel be subject to meltdown during the course of the accident. The German risk analysis study 'Deutsche Risikostudie Kernkraftwerke' dated 1979 considers an explosion of a size sufficient to represent a threat to containment to be considerably more improbable than was assumed in WASH-1400. (orig./DG) [de

Papers presented as part of the status report of the Nuclear Safety Research Project of the Karlsruhe Nuclear Center on 23 March 1994

International Nuclear Information System (INIS)

Hueper, R.

1994-05-01

The ten papers deal with the state of safety requirements on future LWR plants and with nuclear safety research with regard to fast reactors and future PWR plants. In particular, passive after-heat removal, core disruptive accidents, and actinide burning in fast reactors are analysed. For PWR type plants the fuel element behaviour in the event of accidents, hydrogen distribution and hydrogen fires, and the origin and effects of steam explosions on the reactor pressure vessel and the containment are examined. Core meltdown cooling systems are suggested. (DG) [de

Innovations in systems engineering and analysis for the simulation of beyond design-base accidents

International Nuclear Information System (INIS)

Frisch, W.; Beraha, D.

1990-01-01

An important target in improving reactor safety is to have the most realistic simulation possible of beyond design-base accidents in the computer. This paper presents new developments in ATHLET and further developments (description of the thermo-fluid-dynamic conditions in the core and cooling circuits during serious incidents in the computer programme ATHLET-SA) and extensions (link-up to RALOC). RALOC is a computer programme for describing thermodynamic conditions inside the containment during design-base accidents and accidents involving core meltdown. Further research is dedicated to code acceleration. (DG) [de

Analyses of severe accident scenarios in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.; Rimkevicius, S.; Uspuras, E.; Urbonavicius, E.

2006-01-01

Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena. (orig.)

The Chernobyl reactor accident. Pt. 1 and 2

International Nuclear Information System (INIS)

1986-06-01

The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

Performance Test of Core Protection and Monitoring Algorithm with DLL for SMART Simulator Implementation

International Nuclear Information System (INIS)

Koo, Bonseung; Hwang, Daehyun; Kim, Keungkoo

2014-01-01

A multi-purpose best-estimate simulator for SMART is being established, which is intended to be used as a tool to evaluate the impacts of design changes on the safety performance, and to improve and/or optimize the operating procedure of SMART. In keeping with these intentions, a real-time model of the digital core protection and monitoring systems was developed and the real-time performance of the models was verified for various simulation scenarios. In this paper, a performance test of the core protection and monitoring algorithm with a DLL file for the SMART simulator implementation was performed. A DLL file of the simulator application code was made and several real-time evaluation tests were conducted for the steady-state and transient conditions with simulated system variables. A performance test of the core protection and monitoring algorithms for the SMART simulator was performed. A DLL file of the simulator version code was made and several real-time evaluation tests were conducted for various scenarios with a DLL file and simulated system variables. The results of all test cases showed good agreement with the reference results and some features caused by algorithm change were properly reflected to the DLL results. Therefore, it was concluded that the SCOPS S SIM and SCOMS S SIM algorithms and calculational capabilities are appropriate for the core protection and monitoring program in the SMART simulator

Optimizing Decision Preparedness by Adapting Scenario Complexity and Automating Scenario Generation

Science.gov (United States)

Dunne, Rob; Schatz, Sae; Flore, Stephen M.; Nicholson, Denise

2011-01-01

Klein's recognition-primed decision (RPD) framework proposes that experts make decisions by recognizing similarities between current decision situations and previous decision experiences. Unfortunately, military personnel arQ often presented with situations that they have not experienced before. Scenario-based training (S8T) can help mitigate this gap. However, SBT remains a challenging and inefficient training approach. To address these limitations, the authors present an innovative formulation of scenario complexity that contributes to the larger research goal of developing an automated scenario generation system. This system will enable trainees to effectively advance through a variety of increasingly complex decision situations and experiences. By adapting scenario complexities and automating generation, trainees will be provided with a greater variety of appropriately calibrated training events, thus broadening their repositories of experience. Preliminary results from empirical testing (N=24) of the proof-of-concept formula are presented, and future avenues of scenario complexity research are also discussed.

Prediction of LOCA Break Size Using CFNN

Energy Technology Data Exchange (ETDEWEB)

Choi, Geon Pil; Yoo, Kwae Hwan; Back, Ju Hyun; Kim, Dong Yeong; Na, Man Gyun [Chosun University Gwangju (Korea, Republic of)

2016-05-15

The NPPs have the emergency core cooling system (ECCS) such as a safety injection system. The ECCS may not function properly in case of the small break size due to a slight change of pressure in the pipe. If the coolant is not supplied by ECCS, the reactor core will melt. Therefore, the meltdown of reactor core have to be prevented by appropriate accident management through the prediction of LOCA break size in advance. This study presents the prediction of LOCA break size using cascaded fuzzy neural network (CFNN). The CFNN model repeatedly applies FNN modules that are serially connected. The CFNN model is a data-based method that requires data for its development and verification. The data were obtained by numerically simulating severe accident scenarios of the optimized power reactor (OPR1000) using MAAP code, because real severe accident data cannot be obtained from actual NPP accidents. The CFNN model has been designed to rapidly predict the LOCA break size in LOCA situations. The CFNN model was trained by using the training data set and checked by using test data set. These data sets were obtained using MAAP code for OPR1000 reactor. The performance results of the CFNN model show that the RMS error decreases as the stage number of the CFNN model increases. In addition, the performance result of the CFNN model presents that the RMS error level is below 4%.

Special national report of the Slovak Republic compiled under the convention on nuclear safety. April 2012

International Nuclear Information System (INIS)

2012-04-01

A Special safety report of the Slovak Republic in 2012 is presented. An account of activities carried out by the Nuclear Regulatory Authority of the Slovak Republic (UJD) is presented. These activities are reported under the headings: (0) Introduction; (0.1) Purpose of the report; (0.2) Brief description of the site characteristics and units; (1) Executive summary; (2) External events; (2.1) Seismic; (2.2) Flooding; (2.3) Extreme weather conditions; (3) Design issues; (3.1) Loss of electrical power; (3.2) Loss of the decay heat removal capability/ultimate heat sink; (3.3) Loss of the primary ultimate heat sink, combined with station black out (see stress tests specifications); (4) Severe accident management; (4.1) Organization and arrangements of the licensee to manage accidents; (4.2) Accident management measures in place at the various stages of a scenario of loss of the core cooling function; (4.3) Maintaining the containment integrity after occurrence of significant fuel damage (up to core meltdown) in the reactor core; (4.4) Accident management measures to restrict the radioactive releases; (5) National organizations (regulator, technical support organizations, operator, government); (5.1) Legislative and regulatory framework; (6) Emergency preparedness and response and post--accident management (off-site); (6.1) Implementation of legislation in the field of emergency preparedness; (7) International cooperation; (7.1) Conventions and communications; (7.2) Cooperation with the international organizations; (7.3) Providing feedback including occurrences at nuclear installations of other nuclear power plants abroad.

Safety analysis for push-mode and rotary-mode core sampling

International Nuclear Information System (INIS)

Milliken, N.J.; Geschke, G.R.

1995-01-01

This safety analysis analyzes using the push-mode core sampling truck in the push-mode and the rotary-mode core sampling trucks in both the push- and rotary-modes to retrieve core samples that, once taken and analyzed, will yield waste characterization data for the hazardous waste tanks at the Hanford Site. Operation of the core sampling trucks in both the push- and rotary-modes was reviewed to determine whether the release of radioactive materials could occur during operation. It was concluded that there are three credible scenarios: a sample spill outside of the tank, a steam release event, and an unfiltered release to the environment during continuous exhauster operation. The probability of a sample spill was found to be 10 -4 /event, the probability of a steam release event was determined to fall in the unlikely range (10 -2 /event to 10 -4 /event), and the probability of an unfiltered release was calculated to be 5 x 10 -3 /year. Typically, events with probabilities of 10 -6 /event or less are not considered to be risk significant, and the consequences usually are not analyzed. The three accident scenarios were analyzed to calculate the dose consequences. It was determined that the steam release event is the bounding accident. The onsite and offsite dose consequences for this event are calculated to be 0.24 Sv (24 rem) and 3.2 x 10 -4 Sv (32 mrem), respectively. These consequences are below the risk acceptance guidelines for an unlikely event, as established in WHC-CM-4-46, Nonreactor Facility Safety Analysis Manual. With the design features and the use of the controls presented in Section 8.0, this operation represents a minimal risk

RBMK-1500 accident management for loss of long-term core cooling

International Nuclear Information System (INIS)

Uspuras, E.; Kaliatka, A.

2001-01-01

Results of the Level 1 probabilistic safety assessment of the Ignalina NPP has shown that in topography of the risk, transients dominate above the accidents with LOCAs and failure of the core long-term cooling are the main factors to frequency of the core damage. Previous analyses have shown, that after initial event, as a rule, the reactivity control, as well as short-term and intermediate cooling are provided. However, the acceptance criteria of the long-term cooling are not always carried out. It means that from this point of view the most dangerous accident scenarios are the scenarios related to loss of the core long-term cooling. On the other hand, the transition to the core condition due to loss of the long-term cooling specifies potential opportunities for the management of the accident consequences. Hence, accident management for the mitigation of the accident consequences should be considered and developed. The most likely initiating event, which probably leads to the loss of long term cooling accident, is station blackout. The station blackout is the loss of normal electrical power supply for local needs with an additional failure on start-up of all diesel generators. In the case of loss of electrical power supply MCPs, the circulating pumps of the service water system and MFWPs are switched-off. At the same time, TCV of both turbines are closed. Failure of diesel generators leads to the non-operability of the ECCS long-term cooling subsystem. It means the impossibility to feed MCC by water. The analysis of the station blackout for Ignalina NPP was performed using RELAP5 code. (author)

Emissions reduction scenarios in the Argentinean Energy Sector

International Nuclear Information System (INIS)

Di Sbroiavacca, Nicolás; Nadal, Gustavo; Lallana, Francisco; Falzon, James; Calvin, Katherine

2016-01-01

pathways, the models agree that fossil fuels, in particular natural gas, will remain an important part of the electricity mix in the core baseline scenario. According to the models there is agreement that the introduction of a carbon price will lead to a decline in absolute and relative shares of aggregate fossil fuel generation. However, predictions vary as to the extent to which coal, nuclear and renewable energy play a role. - Highlights: • A scenario that incorporates mitigation measures considered most feasible by relevant Argentinean stakeholders generates a CO_2e emissions reduction of 16% compared to BAU. • This scenario has a total additional cumulative cost of $22.8 Billion (2005 USD) along the period 2010–2050 • A high CO_2 price scenario in LEAP generates a CO_2e emissions reduction of 11.3% compared to the baseline • A high CO_2 price scenario in TIAM-ECN and GCAM generates reductions in CO_2e emissions of 37% and 94% respectively. The main reason for this difference between models includes varying assumptions about technology cost and availability, CO_2 storage capacity, and the ability to import bioenergy. • Under climate policy, natural gas remains an important part of the energy mix, and although models agree that aggregate and proportional fossil fuel use declines, predictions vary as to the extent coal, nuclear and renewable energy play a role moving forward.

Classifying Scenarios in a Product Design Process: a study to achieve automated scenario generation

NARCIS (Netherlands)

Anggreeni, Irene; van der Voort, Mascha C.; van Houten, F.J.A.M.; Miedema, J.; Lutters, D.

2008-01-01

This paper explains the possible uses of scenarios in product design. A scenario classification is proposed as a framework to create, use and reuse different types of scenarios in a product design process. Our aims are three-fold: (1) to obtain a better view on the extent to which scenarios can be

Analysis of a postulated accident scenario involving loss of forced flow in a LMFBR

International Nuclear Information System (INIS)

Moreira, M.L.

1985-01-01

A model to analyse a postulated accident scenario involving loss of forced flow in the reactor vessel of a LMFBR is used. Five phases of the accident are analysed: Natural Circulation, Subcooled Boiling, Nucleate Boiling, Core Dryout and Cladding melt. The heat conduction in the fuel cladding, coolant and lower and upper plenum are calculated by a lump-parameter model. Physical data of a prototype LMFBR reactor were used for the calculation. (author)

IPCC workshop on socio-economic scenarios. Workshop report

Energy Technology Data Exchange (ETDEWEB)

Edenhofer, O.; Pichs-Madruga, R.; Sokona, Y. (and others)

2012-07-01

The goal of the IPCC Workshop on Socio-Economic Scenarios (WoSES) was to facilitate the development of socioeconomic narratives and pathways by the integrated assessment modelling, impacts, and adaptation communities. Describing these pathways and narratives is a core step to analyzing the interdependent issues of adaptation and mitigation in an integrated manner. The Workshop participants agreed that structured and consistent assessments of possible future impacts, vulnerabilities, adaptation, and mitigation would benefit from using shared qualitative narrative and quantitative descriptions of potential socioeconomic and ecosystem reference conditions that underlie challenges to mitigation and adaptation. These descriptions should be flexible enough to provide a framework for comparison within which regional or local studies of adaptation and vulnerability could build their own narratives. The defining socioeconomic conditions of these scenarios are designated Shared Socioeconomic reference Pathways (SSPs). The SSPs define the state of human and natural societies at a macro scale and have two elements: a narrative storyline and a set of quantified measures that define the high-level state of society as it evolves over the 21st century under the assumption of no significant climate change. This assumption defines the SSPs as a baseline independent of climate change projections. The set of SSPs was chosen to characterize the range of uncertainty in mitigation efforts required to achieve particular radiative forcing pathways, in adaptation efforts that could be undertaken to prepare for and respond to the climate change associated with those pathways, and in residual impacts. This will allow assessment of scenarios along two axes: socioeconomic challenges to mitigation, and socioeconomic challenges to adaptation. This conceptualization of SSPs allows them to be combined with different degrees of anthropogenic interference with the climate system (measured in terms of

Using SAFRAN Software to Assess Radiological Hazards from Dismantling of Tammuz-2 Reactor Core at Al-tuwaitha Nuclear Site

Science.gov (United States)

Abed Gatea, Mezher; Ahmed, Anwar A.; jundee kadhum, Saad; Ali, Hasan Mohammed; Hussein Muheisn, Abbas

2018-05-01

The Safety Assessment Framework (SAFRAN) software has implemented here for radiological safety analysis; to verify that the dose acceptance criteria and safety goals are met with a high degree of confidence for dismantling of Tammuz-2 reactor core at Al-tuwaitha nuclear site. The activities characterizing, dismantling and packaging were practiced to manage the generated radioactive waste. Dose to the worker was considered an endpoint-scenario while dose to the public has neglected due to that Tammuz-2 facility is located in a restricted zone and 30m berm surrounded Al-tuwaitha site. Safety assessment for dismantling worker endpoint-scenario based on maximum external dose at component position level in the reactor pool and internal dose via airborne activity while, for characterizing and packaging worker endpoints scenarios have been done via external dose only because no evidence for airborne radioactivity hazards outside the reactor pool. The in-situ measurements approved that reactor core components are radiologically activated by Co-60 radioisotope. SAFRAN results showed that the maximum received dose for workers are (1.85, 0.64 and 1.3mSv/y) for activities dismantling, characterizing and packaging of reactor core components respectively. Hence, the radiological hazards remain below the low level hazard and within the acceptable annual dose for workers in radiation field

Feedback controlled, reactor relevant, high-density, high-confinement scenarios at ASDEX Upgrade

Science.gov (United States)

Lang, P. T.; Blanken, T. C.; Dunne, M.; McDermott, R. M.; Wolfrum, E.; Bobkov, V.; Felici, F.; Fischer, R.; Janky, F.; Kallenbach, A.; Kardaun, O.; Kudlacek, O.; Mertens, V.; Mlynek, A.; Ploeckl, B.; Stober, J. K.; Treutterer, W.; Zohm, H.; ASDEX Upgrade Team

2018-03-01

One main programme topic at the ASDEX Upgrade all-metal-wall tokamak is development of a high-density regime with central densities at reactor grade level while retaining high-confinement properties. This required development of appropriate control techniques capable of coping with the pellet tool, a powerful means of fuelling but one which presented challenges to the control system for handling of related perturbations. Real-time density profile control was demonstrated, raising the core density well above the Greenwald density while retaining the edge density in order to avoid confinement losses. Recently, a new model-based approach was implemented that allows direct control of the central density. Investigations focussed first on the N-seeding scenario owing to its proven potential to yield confinement enhancements. Combining pellets and N seeding was found to improve the divertor buffering further and enhance the operational range accessible. For core densities up to about the Greenwald density, a clear improvement with respect to the non-seeding reference was achieved; however, at higher densities this benefit is reduced. This behaviour is attributed to recurrence of an outward shift of the edge density profile, resulting in a reduced peeling-ballooning stability. This is similar to the shift seen during strong gas puffing, which is required to prevent impurity influx in ASDEX Upgrade. First tests indicate that highly-shaped plasma configurations like the ITER base-line scenario, respond very well to pellet injection, showing efficient fuelling with no measurable impact on the edge density profile.

Scenario planning and nanotechnological futures

International Nuclear Information System (INIS)

Farber, Darryl; Lakhtakia, Akhlesh

2009-01-01

Scenario planning may assist us in harnessing the benefits of nanotechnology and managing the associated risks for the good of the society. Scenario planning is a way to describe the present state of the world and develop several hypotheses about the future of the world, thereby enabling discussions about how the world ought to be. Scenario planning thus is not only a tool for learning and foresight, but also for leadership. Informed decision making by experts and political leaders becomes possible, while simultaneously allaying the public's perception of the risks of new and emerging technologies such as nanotechnology. Two scenarios of the societal impact of nanotechnology are the mixed-signals scenario and the confluence scenario. Technoscientists have major roles to play in both scenarios.

Land-Use Scenarios: National-Scale Housing-Density Scenarios Consistent with Climate Change Storylines (Final Report)

Science.gov (United States)

EPA announced the availability of the final report, Land-Use Scenarios: National-Scale Housing-Density Scenarios Consistent with Climate Change Storylines. This report describes the scenarios and models used to generate national-scale housing density scenarios for the con...

Realistic modelling of external flooding scenarios - A multi-disciplinary approach

International Nuclear Information System (INIS)

Brinkman, J.L.

2014-01-01

Extreme phenomena, such as storm surges or high river water levels, may endanger the safety of nuclear power plants (NPPs) by inundation of the plant site with subsequent damage on safety-related buildings. Flooding may result in simultaneous failures of safety-related components, such as service water pumps and electrical equipment. In addition, the accessibility of the plant may be impeded due to flooding of the plant environment. These consequences are so severe that, (re)assessments of flood risk and flood protection measures should be based on accurate state-of-the-art methods. Dutch nuclear regulations require that a nuclear power plant shall withstand all external initiating events with a return period lower than one million years. For external flooding, this requirement is the basis of the so-called nuclear design level (nucleair ontwerp peil, NOP) of the buildings for external flooding, i.e. the water level at which a system - among others, the nuclear island and the ultimate heat sink - should still function properly. In determining the NOP, the mean water level, wave height and wave behaviour during storm surges are taken into account. This concept could also be used to implement external flooding in a PSA, by assuming that floods exceeding NOP levels directly lead to core damage. However, this straightforward modelling ignores some important aspects: the first is the mitigating effect of the external flood protection as dikes or dunes; the second aspect is that although water levels lower than NOP will not directly lead to core damage, they could do so indirectly as a result of combinations of system loss by flooding and random failure of required safety systems that have to bring the plant in a safe, stable state. Time is a third aspect: failure mechanisms need time to develop and time (via duration of the flood) determines the amount of water on site. This paper describes a PSA approach that takes the (structural) reliability of the external defences

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

OpenAIRE

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

2010-01-01

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

Executive summary. Session summaries

International Nuclear Information System (INIS)

2011-01-01

The possibility of stopping and/or delaying the progression of a core melt accident by the use of a recovered water source or by taking benefit of specific engineered systems is taken into account in a number of PSA (Probabilistic Safety Assessment) studies. The likelihood to stop the progression of a core melt-down accident by water injection is generally considered as high in the early phase of core degradation and depends on reactor specific features, nevertheless even in later sequences, e.g. during the relocation in the lower head, cooling still can be achieved but depends on reactor specific features and the accident scenario.. Ongoing, starting and planned experimental programmes address the coolability issues in the different relevant configurations, i.e. reflooding of bundles, debris beds, molten pools and Reactor Pressure Vessel external cooling. There is still a difficulty with present models to predict reliably if reflooding during the early core degradation would or not trigger a cladding oxidation runaway. Whether this is due to deficiencies in thermal-hydraulics description or problems for taking into account the oxidation of melts is a matter of discussion. The code developments are promisingly directed towards a more mechanistic approach using a porous medium modelling able to treat the different configurations of a degraded core. Secondly, the models to describe adequately the relocation of parts of the molten core to the lower head and the debris bed formation still need further development and qualification. Their validation is expected against the results of ongoing experimental programmes. The transposition of results to the reactor scale where multi-dimensional effects are expected needs to be evaluated, all the more as larger scale experiments are probably not feasible

A key to improved ion core confinement in the JET tokamak: ion stiffness mitigation due to combined plasma rotation and low magnetic shear.

Science.gov (United States)

Mantica, P; Angioni, C; Challis, C; Colyer, G; Frassinetti, L; Hawkes, N; Johnson, T; Tsalas, M; deVries, P C; Weiland, J; Baiocchi, B; Beurskens, M N A; Figueiredo, A C A; Giroud, C; Hobirk, J; Joffrin, E; Lerche, E; Naulin, V; Peeters, A G; Salmi, A; Sozzi, C; Strintzi, D; Staebler, G; Tala, T; Van Eester, D; Versloot, T

2011-09-23

New transport experiments on JET indicate that ion stiffness mitigation in the core of a rotating plasma, as described by Mantica et al. [Phys. Rev. Lett. 102, 175002 (2009)] results from the combined effect of high rotational shear and low magnetic shear. The observations have important implications for the understanding of improved ion core confinement in advanced tokamak scenarios. Simulations using quasilinear fluid and gyrofluid models show features of stiffness mitigation, while nonlinear gyrokinetic simulations do not. The JET experiments indicate that advanced tokamak scenarios in future devices will require sufficient rotational shear and the capability of q profile manipulation.

Developing Optimal Procedure of Emergency Outside Cooling Water Injection for APR1400 Extended SBO Scenario Using MARS Code

Energy Technology Data Exchange (ETDEWEB)

Hwang, Jong Rok; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-10-15

In this study, we examined optimum operator actions to mitigate extended SBO using MARS code. Particularly, this paper focuses on analyzing outside core cooling water injection scenario, and aimed to develop optimal extended SBO procedure. Supplying outside emergency cooling water is the key feature of flexible strategy in extended SBO situation. An optimum strategy to maintain core cooling is developed for typical extended SBO. MARS APR1400 best estimate model was used to find optimal procedure. Also RCP seal leakage effect was considered importantly. Recent Fukushima accident shows the importance of mitigation capability against extended SBO scenarios. In Korea, all nuclear power plants incorporated various measures against Fukushima-like events. For APR1400 NPP, outside connectors are installed to inject cooling water using fire trucks or portable pumps. Using these connectors, outside cooling water can be provided to reactor, steam generators (SG), containment spray system, and spent fuel pool. In U. S., similar approach is chosen to provide a diverse and flexible means to prevent fuel damage (core and SFP) in external event conditions resulting in extended loss of AC power and loss of ultimate heat sink. Hence, hardware necessary to cope with extended SBO is already available for APR1400. However, considering the complex and stressful condition encountered by operators during extended SBO, it is important to develop guidelines/procedures to best cope with the event.

A new scenario framework for Climate Change Research: scenario matrix architecture

NARCIS (Netherlands)

van Vuuren, D.P.|info:eu-repo/dai/nl/11522016X; Kriegler, E.; O'Neill, B.C.; Ebi, K.L.; Riahi, K.; Carter, T.R.; Edmonds, J.; Hallegatte, S.; Kram, T.; Mathur, R.; Winkler, H.

2014-01-01

This paper describes the scenario matrix architecture that underlies a framework for developing new scenarios for climate change research. The matrix architecture facilitates addressing key questions related to current climate research and policy-making: identifying the effectiveness of different

Artificial neural network surrogate development of equivalence models for nuclear data uncertainty propagation in scenario studies

Directory of Open Access Journals (Sweden)

Krivtchik Guillaume

2017-01-01

Full Text Available Scenario studies simulate the whole fuel cycle over a period of time, from extraction of natural resources to geological storage. Through the comparison of different reactor fleet evolutions and fuel management options, they constitute a decision-making support. Consequently uncertainty propagation studies, which are necessary to assess the robustness of the studies, are strategic. Among numerous types of physical model in scenario computation that generate uncertainty, the equivalence models, built for calculating fresh fuel enrichment (for instance plutonium content in PWR MOX so as to be representative of nominal fuel behavior, are very important. The equivalence condition is generally formulated in terms of end-of-cycle mean core reactivity. As this results from a physical computation, it is therefore associated with an uncertainty. A state-of-the-art of equivalence models is exposed and discussed. It is shown that the existing equivalent models implemented in scenario codes, such as COSI6, are not suited to uncertainty propagation computation, for the following reasons: (i existing analytical models neglect irradiation, which has a strong impact on the result and its uncertainty; (ii current black-box models are not suited to cross-section perturbations management; and (iii models based on transport and depletion codes are too time-consuming for stochastic uncertainty propagation. A new type of equivalence model based on Artificial Neural Networks (ANN has been developed, constructed with data calculated with neutron transport and depletion codes. The model inputs are the fresh fuel isotopy, the irradiation parameters (burnup, core fractionation, etc., cross-sections perturbations and the equivalence criterion (for instance the core target reactivity in pcm at the end of the irradiation cycle. The model output is the fresh fuel content such that target reactivity is reached at the end of the irradiation cycle. Those models are built and

Misalignment of Magnetic Fields and Outflows in Protostellar Cores

OpenAIRE

Hull, Charles L. H.; Plambeck, Richard L.; Bolatto, Alberto D.; Bower, Geoffrey C.; Carpenter, John M.; Crutcher, Richard M.; Fiege, Jason D.; Franzmann, Erika; Hakobian, Nicholas S.; Heiles, Carl; Houde, Martin; Hughes, A. Meredith; Jameson, Katherine; Kwon, Woojin; Lamb, James W.

2013-01-01

We present results of λ1.3 mm dust-polarization observations toward 16 nearby, low-mass protostars, mapped with ~2."5 resolution at CARMA. The results show that magnetic fields in protostellar cores on scales of ~1000 AU are not tightly aligned with outflows from the protostars. Rather, the data are consistent with scenarios where outflows and magnetic fields are preferentially misaligned (perpendicular), or where they are randomly aligned. If one assumes that outflows emerge along the rotati...

Safety evaluation report in the matter of Offshore Power Systems Floating Nuclear Plants (1-8). Docket No. STN 50-437

International Nuclear Information System (INIS)

1980-02-01

The staff has performed an evaluation of a refractory sacrificial bed, called a core ladle, proposed by Offshore Power Systems (OPS) to be installed in Floating Nuclear Plants (FNPs) to delay the melt-through penetration of molten core debris in the unlikely event of a core meltdown accident. The core ladle design as proposed by OPS is in response to the staff's manufacturing license Condition 4 given in the Final Environmental Statement, Part III, NUREG-0502, dated December 1978 (FES-III). This condition requires that a pad constructed of magnesium oxide, or other equivalent refractory material, be installed in the FNP structure within the lower reactor cavity. The material must provide increased resistance to melt-through by the molten reactor core, must not react with core debris to form a large volume of gases, and must not have any deleterious effects on safety

Heavy impurity confinement in hybrid operation scenario plasmas with a rotating 1/1 continuous mode

Science.gov (United States)

Raghunathan, M.; Graves, J. P.; Nicolas, T.; Cooper, W. A.; Garbet, X.; Pfefferlé, D.

2017-12-01

In future tokamaks like ITER with tungsten walls, it is imperative to control tungsten accumulation in the core of operational plasmas, especially since tungsten accumulation can lead to radiative collapse and disruption. We investigate the behavior of tungsten trace impurities in a JET-like hybrid scenario with both axisymmetric and saturated 1/1 ideal helical core in the presence of strong plasma rotation. For this purpose, we obtain the equilibria from VMEC and use VENUS-LEVIS, a guiding-center orbit-following code, to follow heavy impurity particles. In this work, VENUS-LEVIS has been modified to account for strong plasma flows with associated neoclassical effects arising from such flows. We find that the combination of helical core and plasma rotation augments the standard neoclassical inward pinch compared to axisymmetry, and leads to a strong inward pinch of impurities towards the magnetic axis despite the strong outward diffusion provided by the centrifugal force, as frequently observed in experiments.

Strategic Scenario Construction Made Easy

DEFF Research Database (Denmark)

Duus, Henrik Johannsen

2016-01-01

insights from the area of strategic forecasting (of which scenario planning is a proper subset) and experiences gained from a recent course in that area to develop a simpler, more direct, hands-on method for scenario construction and to provide several ideas for scenario construction that can be used......Scenario planning is a well-known way to develop corporate strategy by creating multiple images of alternative futures. Yet although scenario planning grew from very hands-on strategy development efforts in the military and from operations research dedicated to solving practical problems, the use...... of scenarios in business has, in many cases, remained a cumbersome affair. Very often a large group of consultants, employees and staff is involved in the development of scenarios and strategies, thus making the whole process expensive in terms of time, money and human resources. In response, this article uses...

Resilience of Core-Periphery Networks in the Case of Rich-Club

Directory of Open Access Journals (Sweden)

Matteo Cinelli

2017-01-01

Full Text Available Core-periphery networks are structures that present a set of central and densely connected nodes, namely, the core, and a set of noncentral and sparsely connected nodes, namely, the periphery. The rich-club refers to a set in which the highest degree nodes show a high density of connections. Thus, a network that displays a rich-club can be interpreted as a core-periphery network in which the core is made up of a number of hubs. In this paper, we test the resilience of networks showing a progressively denser rich-club and we observe how this structure is able to affect the network measures in terms of both cohesion and efficiency in information flow. Additionally, we consider the case in which, instead of making the core denser, we add links to the periphery. These two procedures of core and periphery thickening delineate a decision process in the placement of new links and allow us to conduct a scenario analysis that can be helpful in the comprehension and supervision of complex networks under the resilience perspective. The advantages of the two procedures, as well as their implications, are discussed in relation to both network efficiency and node heterogeneity.

Scenario? Guilty!

DEFF Research Database (Denmark)

Kyng, Morten

1992-01-01

Robert Campbell categorizes the word "scenario" as a buzzword, identifies four major uses within HCI and suggests that we adopt new terms differentiating these four uses of the word. My first reaction to reading the article was definitely positive, but rereading it gave me enough second thoughts...... to warrant a response. I should probably confess that I searched my latest paper for the word "scenario" and found eight occurrences, none of which fell in the categories described by Campbell....

Acoustic Waves: A Route to Enhance Sodium Fast Reactor Safety

International Nuclear Information System (INIS)

Jeannot, Jean-Philippe; Baque, François; Cavaro, Matthieu; Gastaldi, Olivier; Lhuilier, Christian; Massacret, Nicolas; Moriot, Jérémy; Paumel, Kévin; Vandergaegen, Matthias; Rodriguez, Gilles

2013-01-01

Improvement to prevent core meltdown and to provide a more robust safety demonstration → Safety objectives: - A level of safety at least equivalent to EPR’s level, - Consolidation of the defence-in-depth principle, - A more robust safety demonstration than those of the Phenix and Superphenix reactor. Acoustic techniques: - Low attenuation by the sodium medium - High velocity of US wave (2289 m.s-1 at 550°C) →

Learning from global emissions scenarios

International Nuclear Information System (INIS)

O'Neill, Brian C; Nakicenovic, Nebojsa

2008-01-01

Scenarios of global greenhouse gas emissions have played a key role in climate change analysis for over twenty years. Currently, several research communities are organizing to undertake a new round of scenario development in the lead-up to the Fifth Assessment Report of the Intergovernmental Panel on Climate Change (IPCC). To help inform this process, we assess a number of past efforts to develop and learn from sets of global greenhouse gas emissions scenarios. We conclude that while emissions scenario exercises have likely had substantial benefits for participating modeling teams and produced insights from individual models, learning from the exercises taken as a whole has been more limited. Model comparison exercises have typically focused on the production of large numbers of scenarios while investing little in assessing the results or the production process, perhaps on the assumption that later assessment efforts could play this role. However, much of this assessment potential remains untapped. Efforts such as scenario-related chapters of IPCC reports have been most informative when they have gone to extra lengths to carry out more specific comparison exercises, but in general these assessments do not have the remit or resources to carry out the kind of detailed analysis of scenario results necessary for drawing the most useful conclusions. We recommend that scenario comparison exercises build-in time and resources for assessing scenario results in more detail at the time when they are produced, that these exercises focus on more specific questions to improve the prospects for learning, and that additional scenario assessments are carried out separately from production exercises. We also discuss the obstacles to better assessment that might exist, and how they might be overcome. Finally, we recommend that future work include much greater emphasis on understanding how scenarios are actually used, as a guide to improving scenario production.

Occult hepatitis B infection: an evolutionary scenario

Directory of Open Access Journals (Sweden)

Lukashov Vladimir V

2008-12-01

Full Text Available Abstract Background Occult or latent hepatitis B virus (HBV infection is defined as infection with detectable HBV DNA and undetectable surface antigen (HBsAg in patients' blood. The cause of an overt HBV infection becoming an occult one is unknown. To gain insight into the mechanism of the development of occult infection, we compared the full-length HBV genome from a blood donor carrying an occult infection (d4 with global genotype D genomes. Results The phylogenetic analysis of polymerase, core and X protein sequences did not distinguish d4 from other genotype D strains. Yet, d4 surface protein formed the evolutionary outgroup relative to all other genotype D strains. Its evolutionary branch was the only one where accumulation of substitutions suggests positive selection (dN/dS = 1.3787. Many of these substitutiions accumulated specifically in regions encoding the core/surface protein interface, as revealed in a 3D-modeled protein complex. We identified a novel RNA splicing event (deleting nucleotides 2986-202 that abolishes surface protein gene expression without affecting polymerase, core and X-protein related functions. Genotype D strains differ in their ability to perform this 2986-202 splicing. Strains prone to 2986-202 splicing constitute a separate clade in a phylogenetic tree of genotype D HBVs. A single substitution (G173T that is associated with clade membership alters the local RNA secondary structure and is proposed to affect splicing efficiency at the 202 acceptor site. Conclusion We propose an evolutionary scenario for occult HBV infection, in which 2986-202 splicing generates intracellular virus particles devoid of surface protein, which subsequently accumulates mutations due to relaxation of coding constraints. Such viruses are deficient of autonomous propagation and cannot leave the host cell until it is lysed.

Analyses of the performance of the ASTRID-like TRU burners in regional scenario studies - 5136

International Nuclear Information System (INIS)

Vezzoni, B.; Gabrielli, F.; Rineiski, A.

2015-01-01

In the past, large Sodium Fast Reactors systems (earlier CAPRA/CADRA, later ESFR and ESFR-like systems) and Accelerator Driven Systems (ADS-EFIT) were considered and extensively studied in Europe for managing MAs/Pu within regional or national scenario studies. After the ASTRID system was proposed in France, ASTRID-like burners could be considered as further options to be investigated. Low conversion ratio (CR) ASTRID-like burner cores (1200 MWth) have been considered at KIT by introducing few modifications with respect to the original French ASTRID design. These modifications allow keeping almost unchanged the main characteristics of the system (e.g. thermal power) and avoiding a strong deterioration of safety parameters (such as sodium void effect) after introduction of large amounts of Pu (more than 20%) and MAs (2-12%) in the fuel. These cores have already been studied at KIT for phase-out scenarios. A constant energy production case, relevant for a European or another regional scenario is considered in the paper. Cases with different shares (from 10 to 30%) of ASTRID-like burners in the nuclear energy fleet are compared. The results show that the ASTRID-like burners allow the use of all TRUs compositions foreseen in the fuel cycle with a proper choice of the MAs to Pu ratios and of the U/TRUs fractions either in phasing-out and on-going nuclear energy utilization conditions. The results show that a mixed fleet composed of 11% burners and 89% ESFR is able to stabilize the MAs in the cycle. The same stabilization is obtained with a fleet composed by 33% burner in combination with LWRs only

Release of gases and their influence on containment integrity during a hypothetical meltdown accident

International Nuclear Information System (INIS)

Hassmann, K.; Reimann, M.

1981-01-01

The sequence of a hypothetical core melt down accident has been subdivided into four phases. Heating up of the core until failure of the core support structure is the first phase. It starts at a certain water level in the reactor pressure vessel (RPV) and ends with the failure of the grid plate. The second phase is characterized by the evaporation of the water in the lower plenum of the RPV. The second phase lasts until a molten core debris is formed. The third phase is concerned with the heating up of the pressure vessel after formation of a molten pool in the lower plenum of the RPV. After pressure vessel failure, the molten corium will interact in the fourth phase with the concrete structure beneath the pressure vessel. In this paper the gas release during all four accident phases and the resulting pressure-time history within the containment of a German standard PWR is given, taking into account violent combustion of hydrogen. In particular, the differences caused by dsestruction of concrete with silicious and with calcareous aggregates has been analyzed. The basis for the results in the 4th phase is the WECHSL code. Long term containment calculations have been performed with the COCMEL-code

Scenarios and innovative systems

International Nuclear Information System (INIS)

2001-11-01

The purpose of this workshop is to present to the GEDEON community the scenarios for the deployment of innovative nuclear solutions. Both steady state situations and possible transitions from the present to new reactors and fuel cycles are considered. Innovative systems that satisfy improved natural resource utilization and waste minimization criteria will be described as well as the R and D orientations of various partners. This document brings together the transparencies of 17 communications given at this workshop: general policy for transmutation and partitioning; Amster: a molten salt reactor (MSR) concept; MSR capabilities; potentials and capabilities of accelerator driven systems (ADS); ADS demonstrator interest as an experimental facility; innovative systems: gas coolant technologies; Pu management in EPR; scenarios with thorium fuel; scenarios at the equilibrium state; scenarios for transition; partitioning and specific conditioning; management of separated radio-toxic elements; European programs; DOE/AAA (Advanced Accelerator Applications) program; OECD scenario studies; CEA research programs and orientations; partitioning and transmutation: an industrial point of view. (J.S.)

Scenarios for Gluino Coannihilation

CERN Document Server

Ellis, John; Luo, Feng; Olive, Keith A

2016-01-01

We study supersymmetric scenarios in which the gluino is the next-to-lightest supersymmetric particle (NLSP), with a mass sufficiently close to that of the lightest supersymmetric particle (LSP) that gluino coannihilation becomes important. One of these scenarios is the MSSM with soft supersymmetry-breaking squark and slepton masses that are universal at an input GUT renormalization scale, but with non-universal gaugino masses. The other scenario is an extension of the MSSM to include vector-like supermultiplets. In both scenarios, we identify the regions of parameter space where gluino coannihilation is important, and discuss their relations to other regions of parameter space where other mechanisms bring the dark matter density into the range allowed by cosmology. In the case of the non-universal MSSM scenario, we find that the allowed range of parameter space is constrained by the requirement of electroweak symmetry breaking, the avoidance of a charged LSP and the measured mass of the Higgs boson, in parti...

Heat transfer analysis to investigate the core catcher plate assembly in SFR

International Nuclear Information System (INIS)

Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

Thermal and stress analyses of meltdown cups for LMFBR safety experiments using SLSF in-reactor loops

Energy Technology Data Exchange (ETDEWEB)

Blomquist, C. A. [Argonne National Lab., IL (United States); Ariman, T. [Univ. of Notre Dame, IN (United States); Pierce, R. D.; Pedersen, D. R. [Argonne National Lab., IL (United States)

1977-07-01

The test trains for the Sodium Loop Safety Facility (SLSF) in-reactor experiments, which simulate hypothetical LMFBR accidents, have a meltdown cup to protect the primary containment from the effects of molten materials. Thermal and stress analyses were performed on the cup which is designed to contain 3.6 kg of molten fuel and 2.4 kg of molten steel. The cup principal components are: 1. A 38 mm diameter tungsten spike which provides initial fuel quenching and prevents fuel boiling, 2. A 73 mm inside diameter tungsten liner to isolate the support vessel from the molten material high initial temperature, 3. An insulator which is an expedient for extending the experiment time, and 4. An Inconel 625 vessel which provides the structural support to withstand the thermal and pressure stresses. The spike, liner, and insulator are supported by a hemispherical tungsten end cap which fits inside the hemispherical bottom of the support vessel. This vessel is attached to the 316 stainless steel test train with an Inconel 750 wire-formed retaining ring. Thermal analyses were performed with the Argonne-modified version of the general heat transfer code THTB, based on the instantaneous addition of 3200/sup 0/K molten fuel with a decay heat of 9 W/gm and 1920/sup 0/K molten steel. These analyses have shown that the cup will adequately cool the molten materials. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The maximum temperature occurs at the center of the fuel region but it is always less than the fuel boiling point. The most severe heating occurs when there is no sodium flow outside the cup. For this case the sodium boils (approximately 1200/sup 0/K) and the Inconel vessel and tungsten liner temperatures are approximately 1250/sup 0/K and 2420/sup 0/K, respectively.

Teaching of severe accident of Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power

International Nuclear Information System (INIS)

Saito, Shinzo

2011-01-01

The Great East Japan Earthquake and accompanied tsunami brought about the severe accident at Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power Co., Inc. For 'No more Fukushima', twelve teaching of the accident was pointed out as follows: 1) natural disasters and external events shall be taken into consideration, 2) severe accident shall be included into safety regulation, 3) all possibility of hydrogen explosion shall be excluded, 4) diversity of safety important component and equipment shall be added with sufficient period of outage, 5) siting of multiple units at the same site shall be avoided at quake-prone country like Japan, 6) accident response environment for operators shall be improved, 7) accident convergence termination system shall be established so as to concentrate technical experience and knowledge, 8) off-site center shall be improved, 9) resident evacuation, consumption limit of food, radiation exposure and soil contamination limit shall be decided openly, 10) nuclear regulation and prevention of disaster shall be conducted by unitary organization to gain public trust, 11) fostering of safety culture among relevant enterprises shall be more encouraged and 12) nuclear industry shall develop reactor such as with no core meltdown or no evacuation and environmental contamination even if reactor core would be meltdown. (T. Tanaka)

High Level Analysis, Design and Validation of Distributed Mobile Systems with CoreASM

Science.gov (United States)

Farahbod, R.; Glässer, U.; Jackson, P. J.; Vajihollahi, M.

System design is a creative activity calling for abstract models that facilitate reasoning about the key system attributes (desired requirements and resulting properties) so as to ensure these attributes are properly established prior to actually building a system. We explore here the practical side of using the abstract state machine (ASM) formalism in combination with the CoreASM open source tool environment for high-level design and experimental validation of complex distributed systems. Emphasizing the early phases of the design process, a guiding principle is to support freedom of experimentation by minimizing the need for encoding. CoreASM has been developed and tested building on a broad scope of applications, spanning computational criminology, maritime surveillance and situation analysis. We critically reexamine here the CoreASM project in light of three different application scenarios.

A working definition of scenario and a method of scenario construction

International Nuclear Information System (INIS)

Barr, G.E.; Dunn, E.

1992-01-01

The event-tree method of scenario construction has been chosen for the Yucca Mountain performance assessment. Its applicability and suitability to the problem are discussed and compared with those of the Nuclear Regulatory Commission (NRC) method. The event-tree method is appropriate for an incompletely characterized site, where there must be an evolving understanding, over time, of the processes at work, for a site that may require analysis of details in specific context, and when the scenario functions to guide site characterization. Anticipating the eventual requirement for using the NRC method, we show that the event-tree method can be translated to the NRC format after final scenario screening

Constraints on The Coupled Thermal Evolution of the Earth's Core and Mantle, The Age of The Inner Core, And The Origin of the 186Os/188Os Core(?) Signal in Plume-Derived Lavas

Science.gov (United States)

Lassiter, J. C.

2005-12-01

-14 TW. In the absence of heat-producing elements in the core, such high heat flow rates require an inner core younger than ~1 Ga and preclude the development of significant 186Os enrichment in the outer core. Experimental studies suggest that potassium may partition into Fe-S-O liquids during core formation. Radioactive decay of potassium in the core could provide an additional heat source and reconcile geophysical evidence for high core/mantle heat flow with apparent geochemical evidence for an ancient inner core. However, high concentrations of chalcophile elements such as Cu in the mantle are inconsistent with significant segregation of a S-rich liquid during core formation, precluding K partitioning into the core by this mechanism. Furthermore, core formation scenarios that would lead to high K content in the core (e.g., core formation prior to terrestrial volatile depletion) also result in high core Pb concentrations. Core/mantle interaction would then produce strong negative correlations between 186Os/188Os and 207Pb/204Pb ratios, but such correlations are not observed. In summary, elevated 186Os/188Os ratios in some plume-derived lavas are unlikely to reflect core/mantle interaction because the inner core is too young for this isotopic signature to have developed in the outer core. Melt generation from pyroxenite or fractionation of PGEs between sulfide melts and monosulfide solid solutions provide alternative mechanisms for generating ancient mantle reservoirs with elevated Pt/Os and 186Os/188Os.

Learning Through Scenario Planning

DEFF Research Database (Denmark)

Balarezo, Jose

level variables, this research corrects this void by investigating the dynamics of organizational learning through the lenses of a corporate scenario planning process. This enhances our scientific understanding of the role that scenario planning might play in the context of organizational learning......This project investigates the uses and effects of scenario planning in companies operating in highly uncertain and dynamic environments. Whereas previous research on scenario planning has fallen short of providing sufficient evidence of its mechanisms and effects on individual or organizational...... and strategic renewal. Empirical evidence of the various difficulties that learning flows has to overcome as it journeys through organizational and hierarchical levels are presented. Despite various cognitive and social psychological barriers identified along the way, the results show the novel...

Radiological consequence analysis for upgradation of Pakistan Research Reactor-1 from 9 to 10 MW

International Nuclear Information System (INIS)

Khan, L.A.; Raza, S.S.

1993-12-01

Radiological consequence analysis has been carried out for upgradation of PARR-I from 9 to 10 MW. A hypothetical loss of coolant accident resulting in core meltdown and release of fission products to the atmosphere has been analyzed. Whole body and thyroid doses have been calculated as a function of time and distance from the containment building. Based on these dose estimates, boundaries of exclusion and low population zones are assessed. (author)

CORCON-MOD3: An integrated computer model for analysis of molten core-concrete interactions

International Nuclear Information System (INIS)

Bradley, D.R.; Gardner, D.R.; Brockmann, J.E.; Griffith, R.O.

1993-10-01

The CORCON-Mod3 computer code was developed to mechanistically model the important core-concrete interaction phenomena, including those phenomena relevant to the assessment of containment failure and radionuclide release. The code can be applied to a wide range of severe accident scenarios and reactor plants. The code represents the current state of the art for simulating core debris interactions with concrete. This document comprises the user's manual and gives a brief description of the models and the assumptions and limitations in the code. Also discussed are the input parameters and the code output. Two sample problems are also given

Scenario-based strategizing

DEFF Research Database (Denmark)

Lehr, Thomas; Lorenz, Ullrich; Willert, Markus

2017-01-01

For over 40 years, scenarios have been promoted as a key technique for forming strategies in uncertain en- vironments. However, many challenges remain. In this article, we discuss a novel approach designed to increase the applicability of scenario-based strategizing in top management teams. Drawi...... Ministry) and a firm affected by disruptive change (Bosch, leading global supplier of technology and solutions)....

Shared memory and message passing revisited in the many-core era

CERN Multimedia

CERN. Geneva

2016-01-01

In the 70s, Edsgar Dijkstra, Per Brinch Hansen and C.A.R Hoare introduced the fundamental concepts for concurrent computing. It was clear that concrete communication mechanisms were required in order to achieve effective concurrency. Whether you're developing a multithreaded program running on a single node, or a distributed system spanning over hundreds of thousands cores, the choice of the communication mechanism for your system must be done intelligently because of the implicit programmability, performance and scalability trade-offs. With the emergence of many-core computing architectures many assumptions may not be true anymore. In this talk we will try to provide insight on the characteristics of these communication models by providing basic theoretical background and then focus on concrete practical examples based on indicative use case scenarios. The case studies of this presentation cover popular programming models, operating systems and concurrency frameworks in the context of many-core processors.

40 Years of Shell Scenarios

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-02-15

Shell has been using scenario planning for four decades. During that time these scenarios have helped the company and governments across the world to make better strategic choices. Scenarios provide lenses that help see future prospects more clearly, make richer judgments and be more sensitive to uncertainties. Discover how the Shell Scenarios team has helped guide decision makers at major moments in history and get a peek at the team future focus, including the intricate relationship between energy, water and food.

Scenarios in society, society in scenarios: toward a social scientific analysis of storyline-driven environmental modeling

International Nuclear Information System (INIS)

Garb, Yaakov; Pulver, Simone; VanDeveer, Stacy D

2008-01-01

Scenario analysis, an approach to thinking about alternative futures based on storyline-driven modeling, has become increasingly common and important in attempts to understand and respond to the impacts of human activities on natural systems at a variety of scales. The construction of scenarios is a fundamentally social activity, yet social scientific perspectives have rarely been brought to bear on it. Indeed, there is a growing imbalance between the increasing technical sophistication of the modeling elements of scenarios and the continued simplicity of our understanding of the social origins, linkages, and implications of the narratives to which they are coupled. Drawing on conceptual and methodological tools from science and technology studies, sociology and political science, we offer an overview of what a social scientific analysis of scenarios might include. In particular, we explore both how scenarios intervene in social microscale and macroscale contexts and how aspects of such contexts are embedded in scenarios, often implicitly. Analyzing the social 'work' of scenarios (i) can enhance the understanding of scenario developers and modeling practitioners of the knowledge production processes in which they participate and (ii) can improve the utility of scenario products as decision-support tools to actual, rather than imagined, decision-makers.

Scenarios for remote gas production

International Nuclear Information System (INIS)

Tangen, Grethe; Molnvik, Mona J.

2009-01-01

The amount of natural gas resources accessible via proven production technology and existing infrastructure is declining. Therefore, smaller and less accessible gas fields are considered for commercial exploitation. The research project Enabling production of remote gas builds knowledge and technology aiming at developing competitive remote gas production based on floating LNG and chemical gas conversion. In this project, scenarios are used as basis for directing research related to topics that affect the overall design and operation of such plants. Selected research areas are safety, environment, power supply, operability and control. The paper summarises the scenario building process as a common effort among research institutes and industry. Further, it documents four scenarios for production of remote gas and outlines how the scenarios are applied to establish research strategies and adequate plans in a multidisciplinary project. To ensure relevance of the scenarios, it is important to adapt the building process to the current problem and the scenarios should be developed with extensive participation of key personnel.

2015 Standard Scenarios Annual Report: U.S. Electric Sector Scenario Exploration

Energy Technology Data Exchange (ETDEWEB)

Sullivan, Patrick [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Cole, Wesley [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Blair, Nate [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Lantz, Eric [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Krishnan, Venkat [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Mai, Trieu [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Mulcahy, David [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Porro, Gian [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

2015-07-16

This report is one of several products resulting from an initial effort to provide a consistent set of technology cost and performance data and to define a conceptual and consistent scenario framework that can be used in the National Renewable Energy Laboratory’s (NREL’s) future analyses. The long-term objective of this effort is to identify a range of possible futures of the U.S. electricity sector in which to consider specific energy system issues through (1) defining a set of prospective scenarios that bound ranges of key technology, market, and policy assumptions and (2) assessing these scenarios in NREL’s market models to understand the range of resulting outcomes, including energy technology deployment and production, energy prices, and carbon dioxide (CO2) emissions.

Scenario Development for Water Resources Planning and Management

Science.gov (United States)

Stewart, S.; Mahmoud, M.; Liu, Y.; Hartman, H.; Wagener, T.; Gupta, H.

2006-12-01

The main objective of scenario development for water resources is to inform policy-makers about the implications of various policies to inform decision-making. Although there have been a number of studies conducted in the relatively-new and recent field of scenario analysis and development, very few of those have been explicitly applied to water resource issues. More evident is the absence of an established formal approach to develop and apply scenarios. Scenario development is a process that evaluates possible future states of the world by examining several feasible scenarios. A scenario is a projection of various physical and socioeconomic conditions that describe change from the current state to a future state. In this paper, a general framework for scenario development with special emphasis on applications to water resources is considered. The process comprises several progressive and reiterative phases: scenario definition, scenario construction, scenario analysis, scenario assessment, and risk management. Several characteristics of scenarios that are important in describing scenarios are also taken into account; these include scenario types, scenario themes, scenario likelihoods and scenario categories. A hindrance to the adoption of a unified framework for scenario development is inconsistency in the terminology used by scenario developers. To address this problem, we propose a consistent terminology of basic and frequent terms. Outreach for this formal approach is partially maintained through an interactive community website that seeks to educate potential scenario developers about the scenario development process, share and exchange information and resources on scenarios to foster a multidisciplinary community of scenario developers, and establish a unified framework for scenario development with regards to terminology and guidelines. The website provides information on scenario development, current scenario-related activities, key water resources scenario

Exposure scenarios for workers

NARCIS (Netherlands)

Marquart, H.; Northage, C.; Money, C.

2007-01-01

The new European chemicals legislation REACH (Registration, Evaluation, Authorisation and restriction of Chemicals) requires the development of Exposure Scenarios describing the conditions and risk management measures needed for the safe use of chemicals. Such Exposure Scenarios should integrate

Assessing climate change impacts, benefits of mitigation, and uncertainties on major global forest regions under multiple socioeconomic and emissions scenarios

Science.gov (United States)

John B Kim; Erwan Monier; Brent Sohngen; G Stephen Pitts; Ray Drapek; James McFarland; Sara Ohrel; Jefferson Cole

2016-01-01

We analyze a set of simulations to assess the impact of climate change on global forests where MC2 dynamic global vegetation model (DGVM) was run with climate simulations from the MIT Integrated Global System Model-Community Atmosphere Model (IGSM-CAM) modeling framework. The core study relies on an ensemble of climate simulations under two emissions scenarios: a...

Regional climate change scenarios

International Nuclear Information System (INIS)

Somot, S.

2005-01-01

Because studies of the regional impact of climate change need higher spatial resolution than that obtained in standard global climate change scenarios, developing regional scenarios from models is a crucial goal for the climate modelling community. The zoom capacity of ARPEGE-Climat, the Meteo-France climate model, allows use of scenarios with a horizontal resolution of about 50 km over France and the Mediterranean basin. An IPCC-A2 scenario for the end of the 21. century in France shows higher temperatures in each season and more winter and less summer precipitation than now. Tuning the modelled statistical distributions to observed temperature and precipitation allows us to study changes in the frequency of extreme events between today's climate and that at the end of century. The frequency of very hot days in summer will increase. In particular, the frequency of days with a maximum temperature above 35 deg C will be multiplied by a factor of 10, on average. In our scenario, the Toulouse area and Provence might see one quarter of their summer days with a maximum temperature above 35 deg C. (author)

Nuclear power prospects and potential: scenarios

International Nuclear Information System (INIS)

Rogner, Hans-Hogler; McDonald, Alan; )

2002-01-01

This paper outlines a range of scenarios describing what the world's energy system might look in the middle of the century, and what nuclear energy's most profitable role might be. The starting point is the 40 non-greenhouse-gas-mitigation scenarios in the Special Report on Emission Scenarios (SRES) of the Intergovernmental Panel on Climate Change (IPCC). Given their international authorship and comprehensive review by governments and scientific experts, the SRES scenarios are the state of the art in long-term energy scenarios

Scenario-based roadmapping assessing nuclear technology development paths for future nuclear energy system scenarios

International Nuclear Information System (INIS)

Van Den Durpel, Luc; Roelofs, Ferry; Yacout, Abdellatif

2009-01-01

Nuclear energy may play a significant role in a future sustainable energy mix. The transition from today's nuclear energy system towards a future more sustainable nuclear energy system will be dictated by technology availability, energy market competitiveness and capability to achieve sustainability through the nuclear fuel cycle. Various scenarios have been investigated worldwide each with a diverse set of assumptions on the timing and characteristics of new nuclear energy systems. Scenario-based roadmapping combines the dynamic scenario-analysis of nuclear energy systems' futures with the technology roadmap information published and analysed in various technology assessment reports though integrated within the nuclear technology roadmap Nuclear-Roadmap.net. The advantages of this combination is to allow mutual improvement of scenario analysis and nuclear technology roadmapping providing a higher degree of confidence in the assessment of nuclear energy system futures. This paper provides a description of scenario-based roadmapping based on DANESS and Nuclear-Roadmap.net. (author)

Elaborating SRES scenarios for nuclear energy

International Nuclear Information System (INIS)

McDonald, Alan; Riahi, Keywan; Rogner, Hans-Holger

2003-01-01

The objective of this paper is identifying mid-century economic targets for nuclear energy. The first step is to describe what the mid-century energy market might look like: the major competitors for nuclear energy, what products are in demand, how much of each, where is growth greatest, and so forth. The mechanism for systematically describing the future market is scenario building. The starting point is the scenarios in the Special Report on Emissions Scenarios (SRES) of the Intergovernmental Panel on Climate Change. SRES developed four narrative story lines, each representing a different coherent set of demographic, social, economic, technological, and environmental developments. For each story line several different scenarios were developed by six international modelling teams, resulting in 40 scenarios grouped in the 4 story lines. For three of the story lines this paper uses a single marker scenario representative of central tendencies within the scenario family. For the fourth story line the authors chose the scenario that assumes that advances in non-fossil technologies - renewable, nuclear, and high-efficiency conservation technologies - make them most cost-competitive. (BA)

HIERARCHICAL GRAVITATIONAL FRAGMENTATION. I. COLLAPSING CORES WITHIN COLLAPSING CLOUDS

Energy Technology Data Exchange (ETDEWEB)

Naranjo-Romero, Raúl; Vázquez-Semadeni, Enrique; Loughnane, Robert M. [Instituto de Radioastronomía y Astrofísica, Universidad Nacional Autónoma de México, Apdo. Postal 3-72, Morelia, Michoacán, 58089, México (Mexico)

2015-11-20

We investigate the Hierarchical Gravitational Fragmentation scenario through numerical simulations of the prestellar stages of the collapse of a marginally gravitationally unstable isothermal sphere immersed in a strongly gravitationally unstable, uniform background medium. The core developes a Bonnor–Ebert (BE)-like density profile, while at the time of singularity (the protostar) formation the envelope approaches a singular-isothermal-sphere (SIS)-like r{sup −2} density profile. However, these structures are never hydrostatic. In this case, the central flat region is characterized by an infall speed, while the envelope is characterized by a uniform speed. This implies that the hydrostatic SIS initial condition leading to Shu's classical inside-out solution is not expected to occur, and therefore neither should the inside-out solution. Instead, the solution collapses from the outside-in, naturally explaining the observation of extended infall velocities. The core, defined by the radius at which it merges with the background, has a time-variable mass, and evolves along the locus of the ensemble of observed prestellar cores in a plot of M/M{sub BE} versus M, where M is the core's mass and M{sub BE} is the critical BE mass, spanning the range from the “stable” to the “unstable” regimes, even though it is collapsing at all times. We conclude that the presence of an unstable background allows a core to evolve dynamically from the time when it first appears, even when it resembles a pressure-confined, stable BE-sphere. The core can be thought of as a ram-pressure confined BE-sphere, with an increasing mass due to the accretion from the unstable background.

TRU composition changes and their influence on FBR core characteristics in the LWR-to-FBR transition period

International Nuclear Information System (INIS)

Maruyama, Shuhei; Ohki, Shigeo; Mizuno, Tomoyasu

2009-01-01

In the conceptual core and fuel design studies in Fast Reactor Cycle Technology Development Project (FaCT Project) in Japan, much interest has been taken in the fuel nuclide compositions for a transition period from light water reactor to fast breeder reactor (FBR). In this paper, the range of transuranic (TRU) nuclide composition to be provided to FBR is evaluated with extended recycling scenario calculations. The influence of TRU composition changes on FBR core characteristics are also discussed with explanations of major contributing factors. (author)

Energy scenarios for New Zealand

Energy Technology Data Exchange (ETDEWEB)

Harris, G. S.; Ellis, M. J.; Scott, G. C.; Wood, J. R.

1977-10-15

Three energy scenarios have been formulated for New Zealand. They concentrate on those aspects of society which have a direct bearing on energy, emphasizing three important issues: major shifts in society's values in relation to material wealth, pollution, and resources. The scenarios make assumptions that certain overall social conditions would prevail so that all decisions of government, the private sector, and individuals would be governed by the requirement to conform to the scenario theme in a way not possible under existing social and political conditions. The 3 scenarios are known as Continuation, Low New Zealand Pollution, and Limited Growth.

Transient debris freezing and potential wall melting during a severe reactivity initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Moore, R.L.

1981-01-01

It is important to light water reactor (LWR) safety analysis to understand the transient freezing of molten core debris on cold structures following a hypothetical core meltdown accident. The purpose of this paper is to (a) present the results of a severe reactivity initiated accident (RIA) in-pile experiment with regard to molten debris distribution and freezing following test fuel rod failure, (b) analyze the transient freezing of molten debris (primarily a mixture of UO/sub 2/ fuel and Zircaloy cladding) deposited on the inner surface of the test shroud wall upon rod failure, and (c) assess the potential for wall melting upon being contacted by the molten debris. 26 refs

Energy scenarios: a prospective outlook

International Nuclear Information System (INIS)

Salomon, Thierry; Claustre, Raphael; Charru, Madeleine; Sukov, Stephane; Marignac, Yves; Fink, Meike; Bibas, Ruben; Le Saux, Gildas

2011-01-01

A set of articles discusses the use of energy scenarios: how useful they can be to describe a possible future and even to gather the involved actors, how they have been used in France in the past (for planning or prediction purposes, with sometimes some over-assessed or contradictory results, without considering any decline of nuclear energy, or by setting an impossible equation in the case of the Grenelle de l'Environnement), how the scenario framework impacts its content (depending on the approach type: standard, optimization, bottom-up, top-down, or hybrid). It also discusses the issue of choice of hypotheses on growth-based and de-growth-based scenarios, outlines how energy saving is a key for a sustainable evolution. Two German scenarios regarding electricity production (centralisation or decentralisation) and French regional scenarios for Nord-Pas-de-Calais are then briefly discussed

THE CORE COMPOSITION OF A WHITE DWARF IN A CLOSE DOUBLE-DEGENERATE SYSTEM

International Nuclear Information System (INIS)

Vennes, S.; Kawka, A.

2012-01-01

We report the identification of the double-degenerate system NLTT 16249 that comprises a normal, hydrogen-rich (DA) white dwarf and a peculiar, carbon-polluted white dwarf (DQ) showing photospheric traces of nitrogen. We disentangled the observed spectra and constrained the properties of both stellar components. In the evolutionary scenario commonly applied to the sequence of DQ white dwarfs, both carbon and nitrogen would be dredged up from the core. The C/N abundance ratio (≈50) in the atmosphere of this unique DQ white dwarf suggests the presence of unprocessed material ( 14 N) in the core or in the envelope. Helium burning in the DQ progenitor may have terminated early on the red giant branch after a mass-ejection event leaving unprocessed material in the core, although current mass estimates do not favor the presence of a low-mass helium core. Alternatively, some nitrogen in the envelope may have survived an abridged helium-core burning phase prior to climbing the asymptotic giant branch. Based on available data, we estimate a relatively short orbital period (P ∼< 13 hr) and ongoing spectroscopic observations will help determine precise orbital parameters.

Negatep: A Scenario for Combating Global Warming; Le scenario Negatep. Un scenario de lutte contre le rechauffement climatique

Energy Technology Data Exchange (ETDEWEB)

Acket, C.; Bacher, P. [Sauvons Le Climat, 92 - Boulogne Billancourt (France)

2011-07-15

There have been an increasing number of foresight exercises in the field of energy and global warming in recent years, as we have seen from the articles devoted to these questions by Futuribles in 2011 (both in this special issue and in the April number). It is certainly the case that the goals for greenhouse-gas emission reduction are rather ambitious, particularly in France, it being the aim of the 2005 French framework law on energy to reduce carbon gas discharges by a factor of four. Among these scenarios, the Negatep scenario developed by Claude Acket and Pierre Bacher from the 'Sauvons le climat' [Let's save the climate] Association proposes to achieve this ('factor 4') goal in France by 2050 by reducing fossil fuel use by 75% and replacing this as quickly as possible with electricity produced from non-carbon-gas-emitting sources - chiefly, nuclear power and renewables. The authors lay out their goals here, backed up by figures, comparing these with the reference scenario. They also show the path that must be followed to arrive at these goals, particularly in the residential and tertiary sectors, and in transport and industry (through control of needs and recourse to alternative energy sources). They close by comparing the Negatep scenario with two other more recent scenarios aimed also at reducing greenhouse gas emissions, on the one hand in Europe, and on the other in Germany. The comparison confirms that they were right to rely on electricity as a substitute for oil, but gives them cause for concern in respect of the consequences (formidable in their view) that the replacement of nuclear power and coal energy by intermittent renewable energies might have in Europe, both with regard to costs and to the effects on the power network. (authors)

Polymorphic Nature of Iron and Degree of Lattice Preferred Orientation Beneath the Earth's Inner Core Boundary

Science.gov (United States)

Mattesini, Maurizio; Belonoshko, Anatoly B.; Tkalčić, Hrvoje

2018-01-01

Deciphering the polymorphic nature and the degree of iron lattice-preferred orientation in the Earth's inner core holds a key to understanding the present status and evolution of the inner core. A multiphase lattice-preferred orientation pattern is obtained for the top 350 km of the inner core by means of the ab initio based Candy Wrapper Velocity Model coupled to a Monte Carlo phase discrimination scheme. The achieved geographic distribution of lattice alignment is characterized by two regions of freezing, namely within South America and the Western Central Pacific, that exhibit an uncommon high degree of lattice orientation. In contrast, widespread regions of melting of relatively weak lattice ordering permeate the rest of the inner core. The obtained multiphase lattice-preferred orientation pattern is in line with mantle-constrained geodynamo simulations and allows to setup an ad hoc mineral physics scenario for the complex Earth's inner core. It is found that the cubic phase of iron is the dominating iron polymorph in the outermost part of the inner core.

Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

International Nuclear Information System (INIS)

Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

2009-01-01

A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

Engaging Personas and Narrative Scenarios

DEFF Research Database (Denmark)

Nielsen, Lene

2004-01-01

design ideas. The concept of engaging personas and narrative scenario explores personas in the light of what what it is to identify with and have empathy with a character. The concept of narrative scenarios views the narrative as aid for exploration of design ideas. Both concepts incorporate...... a distinktion between creating, writing and reading. Keywords: personas, scenarios, user-centered design, HCI...

Mars base buildup scenarios

International Nuclear Information System (INIS)

Blacic, J.D.

1985-01-01

Two surface base build-up scenarios are presented in order to help visualize the mission and to serve as a basis for trade studies. In the first scenario, direct manned landings on the Martian surface occur early in the missions and scientific investigation is the main driver and rationale. In the second scenario, early development of an infrastructure to exploite the volatile resources of the Martian moons for economic purposes is emphasized. Scientific exploration of the surface is delayed at first, but once begun develops rapidly aided by the presence of a permanently manned orbital station

Environmental impacts of an accident with a nuclear power plant

International Nuclear Information System (INIS)

1985-10-01

A maximum credible reactor accident is considered: all safety systems fail and the reactor core is not cooled anymore. This so-called meltdown accident is discussed for two different weather situations. For these cases, the effects on public health and environment is studied (radioactive clouds, inhalation and deposition of radioactive materials). The radiation doses calculated are compared with standard levels. In so doing, an estimation is made of the measures necessary to reduce unfavourable consequences. (G.J.P.)

Fabrication of uranium carbide/beryllium carbide/graphite experimental-fuel-element specimens

International Nuclear Information System (INIS)

Muenzer, W.A.

1978-01-01

A method has been developed for fabricating uranium carbide/beryllium carbide/graphite fuel-element specimens for reactor-core-meltdown studies. The method involves milling and blending the raw materials and densifying the resulting blend by conventional graphite-die hot-pressing techniques. It can be used to fabricate specimens with good physical integrity and material dispersion, with densities of greater than 90% of the theoretical density, and with a uranium carbide particle size of less than 10 μm

Third generation nuclear plants

Science.gov (United States)

Barré, Bertrand

2012-05-01

After the Chernobyl accident, a new generation of Light Water Reactors has been designed and is being built. Third generation nuclear plants are equipped with dedicated systems to insure that if the worst accident were to occur, i.e. total core meltdown, no matter how low the probability of such occurrence, radioactive releases in the environment would be minimal. This article describes the EPR, representative of this "Generation III" and a few of its competitors on the world market.

Modelling of DEMO core plasma consistent with SOL/divertor simulations for long-pulse scenarios with impurity seeding

International Nuclear Information System (INIS)

Pacher, G.W.; Pacher, H.D.; Janeschitz, G.; Kukushkin, A.S.; Kotov, V.; Reiter, D.

2007-01-01

The integrated core-pedestal-SOL model is applied to the simulation of a typical DEMO operation. Impurity seeding is used to reduce the power load on the divertor to acceptable levels. The influence on long-pulse operation of impurity seeding with various impurities is investigated. DEMO operation at acceptable peak power loads and long-pulse lengths is demonstrated

From scenarios to components

NARCIS (Netherlands)

Fahland, D.

2010-01-01

Scenario-based modeling has evolved as an accepted paradigm for developing complex systems of various kinds. Its main purpose is to ensure that a system provides desired behavior to its users. A scenario is generally understood as a behavioral requirement, denoting a course of actions that shall

The Structure and Dark Halo Core Properties of Dwarf Spheroidal Galaxies

Science.gov (United States)

Burkert, A.

2015-08-01

The structure and dark matter halo core properties of dwarf spheroidal galaxies (dSphs) are investigated. A double-isothermal (DIS) model of an isothermal, non-self-gravitating stellar system embedded in an isothermal dark halo core provides an excellent fit to the various observed stellar surface density distributions. The stellar core scale length a* is sensitive to the central dark matter density ρ0,d. The maximum stellar radius traces the dark halo core radius {r}c,d. The concentration c* of the stellar system, determined by a King profile fit, depends on the ratio of the stellar-to-dark-matter velocity dispersion {σ }*/{σ }d. Simple empirical relationships are derived that allow us to calculate the dark halo core parameters ρ0,d, {r}c,d, and σd given the observable stellar quantities σ*, a*, and c*. The DIS model is applied to the Milky Way’s dSphs. All dSphs closely follow the same universal dark halo scaling relations {ρ }0,d× {r}c,d={75}-45+85 M⊙ pc-2 that characterize the cores of more massive galaxies over a large range in masses. The dark halo core mass is a strong function of core radius, {M}c,d˜ {r}c,d2. Inside a fixed radius of ˜400 pc the total dark matter mass is, however, roughly constant with {M}d=2.6+/- 1.4× {10}7 M⊙, although outliers are expected. The dark halo core densities of the Galaxy’s dSphs are very high, with {ρ }0,d ≈ 0.2 M⊙ pc-3. dSphs should therefore be tidally undisturbed. Evidence for tidal effects might then provide a serious challenge for the CDM scenario.

Building a low carbon scenario for France. How a participatory approach can enhance social and economic acceptability

International Nuclear Information System (INIS)

Bibas, Ruben; Mathy, Sandrine; Fink, Meike

2012-03-01

This publication presents the French case study of the European ENCI-LowCarb research project: Engaging Civil Society in Low Carbon scenarios. The core activity of this project was the development of a methodology for the transparent integration of stakeholders' contributions in the scenario design process to enhance the stakeholders' acceptance of the resulting low carbon pathways. This attempt at integrating acceptability in scenario-making constitutes an important step to distinguish what is technically and economically feasible from what is acceptable. Today, a wide range of published scenarios emphasize the fact that they are built on public consultations or stakeholders' contributions. However, transparency is lacking concerning the methodology relative to how contributions were taken into account and translated into assumptions that can be used by the modeling tool. The project ENCI-LowCarb aimed at exploring this scientific gap. Energy scenarios outline possible low-carbon futures built around assumptions on fossil fuels prices evolution, technological choices and the mechanisms of energy demand and supply, among others. Scenarios are influential tools in political decision-making processes since they shed light on the long-term impacts of today's investment decisions, especially regarding infrastructures. This is why it is crucial that these pathways derive from discussions with main stakeholders. In this report, the French project team (CIRED and RAC-F) has the pleasure to present energy scenarios for France which derive from a collaborative scenario design process including the participation of a wide range of French stakeholders (civil society organizations including trade unions and non-governmental organizations, private companies, banks, statewide and local authorities). Participating stakeholders were asked to define or select acceptable CO 2 emissions mitigation measures. Their contributions were implemented in the technico-economic model Imaclim

Italian energy scenarios comparative evaluations

International Nuclear Information System (INIS)

Contaldi, Mario

2005-01-01

This paper reviews some representative scenarios of the evolution of the Italian primary energy consumption, updated recently. After an overview of the main macroeconomics assumptions the scenario results are cross checked at sectorial level, with a brief discussion of the underlining data and energy intensity trends. The emissions of CO 2 , SO 2 and NO x resulting from the considered scenarios are also reported and discussed [it

A Massive Prestellar Clump Hosting No High-mass Cores

Energy Technology Data Exchange (ETDEWEB)

Sanhueza, Patricio; Lu, Xing; Tatematsu, Ken’ichi [National Astronomical Observatory of Japan, National Institutes of Natural Sciences, 2-21-1 Osawa, Mitaka, Tokyo 181-8588 (Japan); Jackson, James M. [School of Mathematical and Physical Sciences, University of Newcastle, University Drive, Callaghan, NSW 2308 (Australia); Zhang, Qizhou; Stephens, Ian W. [Harvard-Smithsonian Center for Astrophysics, 60 Garden Street, Cambridge, MA 02138 (United States); Guzmán, Andrés E. [Departamento de Astronomía, Universidad de Chile, Camino el Observatorio 1515, Las Condes, Santiago (Chile); Wang, Ke, E-mail: patricio.sanhueza@nao.ac.jp [European Southern Observatory (ESO) Headquarters, Karl-Schwarzschild-Str. 2, D-85748 Garching bei München (Germany)

2017-06-01

The infrared dark cloud (IRDC) G028.23-00.19 hosts a massive (1500 M {sub ⊙}), cold (12 K), and 3.6–70 μ m IR dark clump (MM1) that has the potential to form high-mass stars. We observed this prestellar clump candidate with the Submillimeter Array (∼3.″5 resolution) and Jansky Very Large Array (∼2.″1 resolution) in order to characterize the early stages of high-mass star formation and to constrain theoretical models. Dust emission at 1.3 mm wavelength reveals five cores with masses ≤15 M {sub ⊙}. None of the cores currently have the mass reservoir to form a high-mass star in the prestellar phase. If the MM1 clump will ultimately form high-mass stars, its embedded cores must gather a significant amount of additional mass over time. No molecular outflows are detected in the CO (2-1) and SiO (5-4) transitions, suggesting that the SMA cores are starless. By using the NH{sub 3} (1, 1) line, the velocity dispersion of the gas is determined to be transonic or mildly supersonic (Δ V {sub nt}/Δ V {sub th} ∼ 1.1–1.8). The cores are not highly supersonic as some theories of high-mass star formation predict. The embedded cores are four to seven times more massive than the clump thermal Jeans mass and the most massive core (SMA1) is nine times less massive than the clump turbulent Jeans mass. These values indicate that neither thermal pressure nor turbulent pressure dominates the fragmentation of MM1. The low virial parameters of the cores (0.1–0.5) suggest that they are not in virial equilibrium, unless strong magnetic fields of ∼1–2 mG are present. We discuss high-mass star formation scenarios in a context based on IRDC G028.23-00.19, a study case believed to represent the initial fragmentation of molecular clouds that will form high-mass stars.

Scenarios

NARCIS (Netherlands)

Pérez-Soba, Marta; Maas, Rob

2015-01-01

We cannot predict the future with certainty, but we know that it is influenced by our current actions, and that these in turn are influenced by our expectations. This is why future scenarios have existed from the dawn of civilization and have been used for developing military, political and economic

SCENARIO PLANNING AS LEARNING

Directory of Open Access Journals (Sweden)

Antonio Lourenço Junior

2010-10-01

Full Text Available Scenario Planning has been increasingly used, from its introduction to the decision process as effective tools to test decisions, and improve performance in a dynamic environment (Chermack, 2005. The purpose of this article is to demonstrate the potential of an experimental Scenario Planning Model to mobilize, encourage and add more content to the organization’s decision making process – mainly with respect to Strategic Plans of two governmental institutions, a pharmaceutical company and a technology education foundation.  This study describes the application stages of a hybrid scenario-planning model – herein referred to as Planning as Learning – via action-research, showing the scenarios resulting from the experiment and describes the main results of an assessment of such practice. In order to do that, two well-established Scenario Planning models (Prospective school and Shell’s model were analyzed. They were used as a reference for the proposition and application of an experimental model in the two study objects. A questionnaire was used to assess the technique impact. It was possible to obtain high levels of reliability. In-depth interviews were also conducted with the participants. At the end, the results confirmed the model efficiency as a basis for decision making in the competitive environment in which the two institutions are inserted, also to encourage the learning process as a group, as observed throughout the work.

SMA millimeter observations of hot molecular cores

Energy Technology Data Exchange (ETDEWEB)

Hernández-Hernández, Vicente; Zapata, Luis; Kurtz, Stan [Centro de Radioastronomía y Astrofísica, Universidad Nacional Autónoma de México, Apdo. Postal 3-72 (Xangari), 58090 Morelia, Michoacán (Mexico); Garay, Guido, E-mail: v.hernandez@crya.unam.mx [Departamento de Astronomía, Universidad de Chile, Camino del Observatorio 1515, Las Condes, Santiago (Chile)

2014-05-01

We present Submillimeter Array observations in the 1.3 mm continuum and the CH{sub 3}CN (12 {sub K}-11 {sub K}) line of 17 hot molecular cores associated with young high-mass stars. The angular resolution of the observations ranges from 1.''0 to 4.''0. The continuum observations reveal large (>3500 AU) dusty structures with gas masses from 7 to 375 M {sub ☉}, which probably surround multiple young stars. The CH{sub 3}CN line emission is detected toward all the molecular cores at least up to the K = 6 component and is mostly associated with the emission peaks of the dusty objects. We used the multiple K-components of the CH{sub 3}CN and both the rotational diagram method and a simultaneous synthetic local thermodynamic equilibrium model with the XCLASS program to estimate the temperatures and column densities of the cores. For all sources, we obtained reasonable fits from XCLASS by using a model that combines two components: an extended and warm envelope and a compact hot core of molecular gas, suggesting internal heating by recently formed massive stars. The rotational temperatures lie in the range of 40-132 K and 122-485 K for the extended and compact components, respectively. From the continuum and CH{sub 3}CN results, we infer fractional abundances from 10{sup –9} to 10{sup –7} toward the compact inner components, which increase with the rotational temperature. Our results agree with a chemical scenario in which the CH{sub 3}CN molecule is efficiently formed in the gas phase above 100-300 K, and its abundance increases with temperature.

Toward an integrated ice core chronology using relative and orbital tie-points

Science.gov (United States)

Bazin, L.; Landais, A.; Lemieux-Dudon, B.; Toyé Mahamadou Kele, H.; Blunier, T.; Capron, E.; Chappellaz, J.; Fischer, H.; Leuenberger, M.; Lipenkov, V.; Loutre, M.-F.; Martinerie, P.; Parrenin, F.; Prié, F.; Raynaud, D.; Veres, D.; Wolff, E.

2012-04-01

Precise ice cores chronologies are essential to better understand the mechanisms linking climate change to orbital and greenhouse gases concentration forcing. A tool for ice core dating (DATICE [developed by Lemieux-Dudon et al., 2010] permits to generate a common time-scale integrating relative and absolute dating constraints on different ice cores, using an inverse method. Nevertheless, this method has only been applied for a 4-ice cores scenario and for the 0-50 kyr time period. Here, we present the bases for an extension of this work back to 800 ka using (1) a compilation of published and new relative and orbital tie-points obtained from measurements of air trapped in ice cores and (2) an adaptation of the DATICE inputs to 5 ice cores for the last 800 ka. We first present new measurements of δ18Oatm and δO2/N2 on the Talos Dome and EPICA Dome C (EDC) ice cores with a particular focus on Marine Isotopic Stages (MIS) 5, and 11. Then, we show two tie-points compilations. The first one is based on new and published CH4 and δ18Oatm measurements on 5 ice cores (NorthGRIP, EPICA Dronning Maud Land, EDC, Talos Dome and Vostok) in order to produce a table of relative gas tie-points over the last 400 ka. The second one is based on new and published records of δO2/N2, δ18Oatm and air content to provide a table of orbital tie-points over the last 800 ka. Finally, we integrate the different dating constraints presented above in the DATICE tool adapted to 5 ice cores to cover the last 800 ka and show how these constraints compare with the established gas chronologies of each ice core.

The negaWatt 2011 scenario

International Nuclear Information System (INIS)

2016-03-01

This article presents the approach adopted for the negaWatt scenario and its obtained results. It is based on sobriety (energy savings), on energy efficiency, and on the use of renewable energies. After having outlined the different reasons for an energy transition (increasing energy consumption, critics and risks related to nuclear energy, and high potential of renewable energies), the scenario is presented with its main principles. The scenario identifies possibilities ranging from half to two thirds of energy saving in the different energy consuming sectors. The building sector is presented as a major issue. The transport is described as a sector to be addressed on the long term. The necessary change of the industry sector is highlighted. The agriculture sector is presented as being at the heart of transition. Energy usages are to become sober, efficient and renewable. The scenario is based on a high rate development of renewable energies, while fossil energies are to become marginal, nuclear is to be progressively and reasonably given up, and networks are to become compatible to ensure the scenario success. Thus, the scenario demonstrates the feasibility of a 100 pc sustainable assessment for primary energy, complies with stakes and objectives by 2050. The cost of energy transition is briefly discussed

Rethinking the role of scenarios: Participatory scripting of low-carbon scenarios for France

International Nuclear Information System (INIS)

Mathy, Sandrine; Fink, Meike; Bibas, Ruben

2015-01-01

This article considers the usefulness of low-carbon scenarios in public decision-making. They may be useful as a product-oriented trajectory. The scenarios on the agenda of the 2013 Energy Debate in France belong to this category. But a scenario may also be process-oriented, in the sense that its scripting process helps build consensus and a minimum level of agreement. We have scripted scenarios using a codevelopment method, involving about 40 stakeholders from the private and public sectors, and from the state: NGOs, consumer groups, trade unions, banks and local authorities. They selected policies they considered acceptable for achieving 75% greenhouse gases emission reductions in 2050. These policies were then integrated in the Imaclim-R-France technico-economic simulation model, as part of a high or moderate acceptability scenario. In the first case emissions were cut by between 58% and 72% by 2050; in the second case by between 68% and 81%, depending on the energy price assumptions. All these measures benefited jobs and economic growth, swiftly and durably cutting household spending on energy services. This offers a solid basis for gaining acceptability for low carbon trajectories; the process constitutes also a framework for consolidating collective learning centering on the acceptability of climate policies. - Highlights: • The article develops a ‘process-oriented’ low carbon scenario for France. • Stakeholders define a set of sectoral and fiscal ‘acceptable’ climate policies. • These policies are integrated within a technico-economic model Imaclim-R-France. • Economic impacts and CO 2 emission reductions are computed. •The co-development methodology favors joint production of solutions and shared vision-building

Carbon-constrained scenarios. Final report

International Nuclear Information System (INIS)

2009-05-01

This report provides the results of the study entitled 'Carbon-Constrained Scenarios' that was funded by FONDDRI from 2004 to 2008. The study was achieved in four steps: (i) Investigating the stakes of a strong carbon constraint for the industries participating in the study, not only looking at the internal decarbonization potential of each industry but also exploring the potential shifts of the demand for industrial products. (ii) Developing an hybrid modelling platform based on a tight dialog between the sectoral energy model POLES and the macro-economic model IMACLIM-R, in order to achieve a consistent assessment of the consequences of an economy-wide carbon constraint on energy-intensive industrial sectors, while taking into account technical constraints, barriers to the deployment of new technologies and general economic equilibrium effects. (iii) Producing several scenarios up to 2050 with different sets of hypotheses concerning the driving factors for emissions - in particular the development styles. (iv) Establishing an iterative dialog between researchers and industry representatives on the results of the scenarios so as to improve them, but also to facilitate the understanding and the appropriate use of these results by the industrial partners. This report provides the results of the different scenarios computed in the course of the project. It is a partial synthesis of the work that has been accomplished and of the numerous exchanges that this study has induced between modellers and stakeholders. The first part was written in April 2007 and describes the first reference scenario and the first mitigation scenario designed to achieve stabilization at 450 ppm CO 2 at the end of the 21. century. This scenario has been called 'mimetic' because it has been build on the assumption that the ambitious climate policy would coexist with a progressive convergence of development paths toward the current paradigm of industrialized countries: urban sprawl, general

Room temperature nanojoining of Cu-Ag core-shell nanoparticles and nanowires

International Nuclear Information System (INIS)

Wang, Jiaqi; Shin, Seungha

2017-01-01

Room temperature (T room , 300 K) nanojoining of Ag has been widely employed in fabrication of microelectronic applications where the shapes and structures of microelectronic components must be maintained. In this research, the joining processes of pure Ag nanoparticles (NPs), Cu-Ag core-shell NPs, and nanowires (NWs) are studied using molecular dynamics simulations at T room . The evolution of densification, potential energy, and structural deformation during joining process are analyzed to identify joining mechanisms. Depending on geometry, different joining mechanisms including crystallization-amorphization, reorientation, Shockley partial dislocation are determined. A three-stage joining scenario is observed in both joining process of NPs and NWs. Besides, the Cu core does not participate in all joining processes, however, it enhances the mobility of Ag shell atoms, contributing to a higher densification and bonding strength at T room , compared with pure Ag nanomaterials. The tensile test shows that the nanojoint bears higher rupture strength than the core-shell NW itself. This study deepens understanding in the underlying joining mechanisms and thus nanojoint with desirable thermal, electrical, and mechanical properties could be potentially achieved.

Treatment of EBR-I NaK mixed waste at Argonne National Laboratory and subsequent land disposal at the Idaho National Engineering and Environmental Laboratory

International Nuclear Information System (INIS)

Herrmann, S. D.; Buzzell, J. A.; Holzemer, M. J.

1998-01-01

Sodium/potassium (NaK) liquid metal coolant, contaminated with fission products from the core meltdown of Experimental Breeder Reactor I (EBR-I) and classified as a mixed waste, has been deactivated and converted to a contact-handled, low-level waste at Argonne's Sodium Component Maintenance Shop and land disposed at the Radioactive Waste Management Complex. Treatment of the EBR-I NaK involved converting the sodium and potassium to its respective hydroxide via reaction with air and water, followed by conversion to its respective carbonate via reaction with carbon dioxide. The resultant aqueous carbonate solution was solidified in 55-gallon drums. Challenges in the NaK treatment involved processing a mixed waste which was incompletely characterized and difficult to handle. The NaK was highly radioactive, i.e. up to 4.5 R/hr on contact with the mixed waste drums. In addition, the potential existed for plutonium and toxic characteristic metals to be present in the NaK, resultant from the location of the partial core meltdown of EBR-I in 1955. Moreover, the NaK was susceptible to degradation after more than 40 years of storage in unmonitored conditions. Such degradation raised the possibility of energetic exothermic reactions between the liquid NaK and its crust, which could have consisted of potassium superoxide as well as hydrated sodium/potassium hydroxides

Treatment of EBR-I NaK mixed waste at Argonne National Laboratory and subsequent land disposal at the Idaho National Engineering and Environmental Laboratory.

Energy Technology Data Exchange (ETDEWEB)

Herrmann, S. D.; Buzzell, J. A.; Holzemer, M. J.

1998-02-03

Sodium/potassium (NaK) liquid metal coolant, contaminated with fission products from the core meltdown of Experimental Breeder Reactor I (EBR-I) and classified as a mixed waste, has been deactivated and converted to a contact-handled, low-level waste at Argonne's Sodium Component Maintenance Shop and land disposed at the Radioactive Waste Management Complex. Treatment of the EBR-I NaK involved converting the sodium and potassium to its respective hydroxide via reaction with air and water, followed by conversion to its respective carbonate via reaction with carbon dioxide. The resultant aqueous carbonate solution was solidified in 55-gallon drums. Challenges in the NaK treatment involved processing a mixed waste which was incompletely characterized and difficult to handle. The NaK was highly radioactive, i.e. up to 4.5 R/hr on contact with the mixed waste drums. In addition, the potential existed for plutonium and toxic characteristic metals to be present in the NaK, resultant from the location of the partial core meltdown of EBR-I in 1955. Moreover, the NaK was susceptible to degradation after more than 40 years of storage in unmonitored conditions. Such degradation raised the possibility of energetic exothermic reactions between the liquid NaK and its crust, which could have consisted of potassium superoxide as well as hydrated sodium/potassium hydroxides.

Biomass Scenario Model | Energy Analysis | NREL

Science.gov (United States)

Biomass Scenario Model Biomass Scenario Model The Biomass Scenario Model (BSM) is a unique range of lignocellulosic biomass feedstocks into biofuels. Over the past 25 years, the corn ethanol plant matter (lignocellulosic biomass) to fermentable sugars for the production of fuel ethanol

Medical students' satisfaction with the Applied Basic Clinical Seminar with Scenarios for Students, a novel simulation-based learning method in Greece.

Science.gov (United States)

Pantelidis, Panteleimon; Staikoglou, Nikolaos; Paparoidamis, Georgios; Drosos, Christos; Karamaroudis, Stefanos; Samara, Athina; Keskinis, Christodoulos; Sideris, Michail; Giannakoulas, George; Tsoulfas, Georgios; Karagiannis, Asterios

2016-01-01

The integration of simulation-based learning (SBL) methods holds promise for improving the medical education system in Greece. The Applied Basic Clinical Seminar with Scenarios for Students (ABCS3) is a novel two-day SBL course that was designed by the Scientific Society of Hellenic Medical Students. The ABCS3 targeted undergraduate medical students and consisted of three core components: the case-based lectures, the ABCDE hands-on station, and the simulation-based clinical scenarios. The purpose of this study was to evaluate the general educational environment of the course, as well as the skills and knowledge acquired by the participants. Two sets of questions were distributed to the participants: the Dundee Ready Educational Environment Measure (DREEM) questionnaire and an internally designed feedback questionnaire (InEv). A multiple-choice examination was also distributed prior to the course and following its completion. A total of 176 participants answered the DREEM questionnaire, 56 the InEv, and 60 the MCQs. The overall DREEM score was 144.61 (±28.05) out of 200. Delegates who participated in both the case-based lectures and the interactive scenarios core components scored higher than those who only completed the case-based lecture session (P=0.038). The mean overall feedback score was 4.12 (±0.56) out of 5. Students scored significantly higher on the post-test than on the pre-test (Pmedical students reported positive opinions about their experiences and exhibited improvements in their clinical knowledge and skills.

Energy scenarios for Colombia - Environmental Aspects

International Nuclear Information System (INIS)

Smith, Ricardo A; Vesga A, Daniel R; Boman, Ulf

2000-01-01

The planning unit of the Colombian ministry of energy -UPME -has done an energy scenario project for Colombia with a 20-year horizon (vision year 2020) in this project the scenario methodology was used in a systemic way involving a great number of local and international energy experts. As a result four energy scenarios were designed and in all of them the possible evolution of all energy was analyzed. In this article a description of the used methodology is presented with the developed scenarios. Also a discussion of the long-range future environmental considerations in the energy sector, taking into account the developed scenarios, is presented. Finally some conclusions and recommendations are presented

Scenario Planning as Organizational Intervention

DEFF Research Database (Denmark)

Balarezo, Jose; Nielsen, Bo Bernhard

2017-01-01

existing contributions on scenario planning within a new consolidating framework that includes antecedents, processes, and outcomes. The proposed framework allows for integration of the extant literature on scenario planning from a wide variety of fields, including strategic management, finance, human...... resource management, operations management, and psychology. Findings: This study contributes to research by offering a coherent and consistent framework for understanding scenario planning as a dynamic process. As such, it offers future researchers with a systematic way to ascertain where a particular......Purpose: This paper identifies four areas in need of future research to enhance our theoretical understanding of scenario planning, and sets the basis for future empirical examination of its effects on individual and organizational level outcomes. Design/methodology/approach: This paper organizes...

Social Foundation of Scenario Planning

DEFF Research Database (Denmark)

Rowland, Nicholas James; Spaniol, Matthew Jon

2017-01-01

In this article, the authors establish that models of scenario planning typically involve a series of phases, stages, or steps that imply a sequenced (i.e., linear or chronological) process. Recursive models, in contrast, allow phases to repeat, thus, incorporating iteration. The authors acknowle......In this article, the authors establish that models of scenario planning typically involve a series of phases, stages, or steps that imply a sequenced (i.e., linear or chronological) process. Recursive models, in contrast, allow phases to repeat, thus, incorporating iteration. The authors...... from science and technology studies (STS) on knowledge production, the authors explain transition from one phase to the next and iteration between and within phases based on social negotiation. To this end, the authors examine the interplay between the “scenario development” phase and the “scenario use......” phase of a planning process with a non-governmental organization in Denmark. The upshot for facilitators is practical insight into how transition between phases and phase iteration in scenario planning can be identified, leveraged, and, thus, managed. The upshot for scholars is a related insight...

IPCC Special report on Emissions Scenarios (SRES)

International Nuclear Information System (INIS)

Anon

2001-01-01

This special report on emissions scenarios (SRES) is intended to reflect the most recent trends in driving forces of emissions; population projections economic development, and structural and technological change. It serves as an update to IS92 scenarios developed by IPCC in the early 1990s to illustrate a plausible range of future greenhouse gas emissions. This update is based on a review of the literature and the development of a database of over 400 global and regional scenarios; 190 of these extend from 1900 to 2100 and thus fed into the development of the narrative scenarios and storylines. Based on the literature review, a set of four alternative scenario families, having a total of 40 emission scenarios have been developed. Each scenario family includes a narrative storyline which describes a demographic, social. economic, technological, environmental and policy future. Characteristic features of each of the four families are summarized and a comparison is made between the IS92 and SRES. One of the main conclusions of this recent scenario construction effort is the realization that alternative combinations of main scenario driving forces can lead to similar levels of GHG emissions by the end of the 21st century, and that scenarios with different underlying assumptions can result in very similar climate change

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

International Nuclear Information System (INIS)

1984-11-01

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

Scenarios for gluino coannihilation

Energy Technology Data Exchange (ETDEWEB)

Ellis, John [Theoretical Particle Physics and Cosmology Group, Department of Physics, King’s College London, London, WC2R 2LS United Kingdom (United Kingdom); Theory Division, CERN,Geneva 23, CH-1211 (Switzerland); Evans, Jason L. [School of Physics and Astronomy, University of Minnesota,Minneapolis, MN, 55455 (United States); William I. Fine Theoretical Physics Institute, School of Physics and Astronomy, University of Minnesota, Minneapolis, MN, 55455 (United States); Luo, Feng [Theory Division, CERN,Geneva 23, CH-1211 (Switzerland); Olive, Keith A. [School of Physics and Astronomy, University of Minnesota,Minneapolis, MN, 55455 (United States); William I. Fine Theoretical Physics Institute, School of Physics and Astronomy, University of Minnesota, Minneapolis, MN, 55455 (United States)

2016-02-11

We study supersymmetric scenarios in which the gluino is the next-to-lightest supersymmetric particle (NLSP), with a mass sufficiently close to that of the lightest supersymmetric particle (LSP) that gluino coannihilation becomes important. One of these scenarios is the MSSM with soft supersymmetry-breaking squark and slepton masses that are universal at an input GUT renormalization scale, but with non-universal gaugino masses. The other scenario is an extension of the MSSM to include vector-like supermultiplets. In both scenarios, we identify the regions of parameter space where gluino coannihilation is important, and discuss their relations to other regions of parameter space where other mechanisms bring the dark matter density into the range allowed by cosmology. In the case of the non-universal MSSM scenario, we find that the allowed range of parameter space is constrained by the requirement of electroweak symmetry breaking, the avoidance of a charged LSP and the measured mass of the Higgs boson, in particular, as well as the appearance of other dark matter (co)annihilation processes. Nevertheless, LSP masses m{sub χ}≲8 TeV with the correct dark matter density are quite possible. In the case of pure gravity mediation with additional vector-like supermultiplets, changes to the anomaly-mediated gluino mass and the threshold effects associated with these states can make the gluino almost degenerate with the LSP, and we find a similar upper bound.

Scenario group summary

International Nuclear Information System (INIS)

Thorndike, A.

1976-01-01

A scenario is given for ISABELLE which provides a plausible sequence of events from FY 1980 to 1990. No doubt reality will be quite different. The scenario is based on the construction schedule of the 1976 proposal. Assembly and testing of the accelerator will occur until the end of FY 1983, and the next six years will provide pp interactions for the initial high energy physics research. By 1990 any temporary conditions associated with start-up of ISABELLE should be a thing of the past and all experimental capabilities fully utilized

Scenario group summary

International Nuclear Information System (INIS)

Thorndike, A.

1976-01-01

The scenario is given which provides a plausible sequence of events for ISABELLE from FY 1980 to 1990. No doubt reality will be quite different. The scenario is based on the construction schedule of the 1976 proposal. Assembly and testing of the accelerator will occur until the end of FY 1983, and the next six years will provide pp interactions for the initial high energy physics research. By 1990 any temporary conditions associated with start-up of ISABELLE should be a thing of the past and all experimental capabilities fully utilized

Corium spreading issue; Le corium et son etalement

Energy Technology Data Exchange (ETDEWEB)

Cognet, G.; Brayer, C.; Cranga, M.; Journeau, C.; Laffont, G.; Splinder, B.; Veteau, J.M. [CEA Grenoble, Dept. de Thermohydraulique et de Physique (DPT), 38 (France)

1999-07-01

Safety is one of the major issues for nuclear power plants; its improvement is a constant R and D axis for the CEA. In the event of a highly unlikely core melt-down accident in Light Water Reactors, the Safety Authorities of several EU countries have requested the industries and utilities to consider severe accidents with reactor pressure vessel failure for the design of the next generation of nuclear power plants. The objective is to preserve the integrity of the containment as the main barrier of fission product release to the environment. This can only be achieved if the core melt mixture (called corium, essentially composed of UO{sub 2}, ZrO{sub 2}, Zr, Fe and fission products) is stabilized before it can penetrate the basement. Consequently, various core-catcher concepts are under investigation for future reactors in order to prevent basement erosion, and to stabilize and control the corium within the containment. In particular, in the EPR (European Pressurized Reactor) core-catcher concept, the corium is mixed with a special concrete, and the molten mixture spread over a large multi-layer surface cooled from the bottom; with subsequent cooling by flooding with water. Therefore, melt spreading requires intensive investigation in order to determine and quantify the key phenomena, which govern the spreading. For some years now, the Nuclear Reactor Division of the Atomic Energy Commission (CEA/DRN) has been conducting a large program to improve knowledge on corium behaviour and coolability. This program is based on experimental (with simulant and prototypic materials) and theoretical investigations, which are finally gathered into scenario and mechanistic computer codes. Within this framework, a large part is currently devoted to the study of corium spreading. After a reminder of the general objectives and a description of the DRn approach and facilities, this paper presents the most important results. (authors)

Radioactive waste assessment using 'moderate growth in nuclear electricity generation' scenario

International Nuclear Information System (INIS)

Richardson, J.A.; Goodill, D.R.; Tymons, B.J.

1985-05-01

This report describes an assessment of radioactive waste management arisings from a defined nuclear power generation scenario -Scheme 3. Scheme 3 assumes a moderate growth in nuclear generation scenario with raw waste arisings from 3 main groups: (i) existing and committed commercial reactors; (ii) fuel reprocessing plants; (iii) research, industry and medicine. No decommissioning wastes are considered except for arisings from the final fuel cores from decommissioned reactors. The study uses the SIMULATION2 code which models waste material flows through the system. With a knowledge of the accumulations and average production rates of the raw wastes and their isotopic compositions (or total activities), the rates at which conditioned wastes become available for transportation and disposal are calculated, with specific activity levels. The data bases for the inventory calculations and the assumptions concerning future operation of nuclear facilities were those current in 1983. Both the inventory data and plans for the future of existing nuclear installations have been updated since these calculations were completed. Therefore the results from this assessment do not represent the most up-to-date information available. The report does, however, illustrate the methodology of assessment, and indicates the type of information that can be generated. (author)

A Parameter Study of Large Fast Reactor Nuclear Explosion Accidents

Energy Technology Data Exchange (ETDEWEB)

Wiesel, J R

1969-02-15

An IBM-code EEM (Explosive Excursion Model) has been developed for calculating the energy releases associated with the explosive disassembly of a large fast reactor following a superprompt critical condition. The assumed failure chain of events and the possible core collapse following a fuel meltdown give the input data and initial conditions, the most important of which is the reactivity insertion rate at the moment of the explosive core disassembly. The dependence of the energy releases on the reactivity insertion rate, the Doppler reactivity feedback, the power form factor and the core size have been studied. The model enables a quick estimation of conservative values of the destructive mechanical energy releases following a nuclear explosion and gives suggestions as to how to reduce or even avoid such excursions.

A Parameter Study of Large Fast Reactor Nuclear Explosion Accidents

International Nuclear Information System (INIS)

Wiesel, J.R.

1969-02-01

An IBM-code EEM (Explosive Excursion Model) has been developed for calculating the energy releases associated with the explosive disassembly of a large fast reactor following a superprompt critical condition. The assumed failure chain of events and the possible core collapse following a fuel meltdown give the input data and initial conditions, the most important of which is the reactivity insertion rate at the moment of the explosive core disassembly. The dependence of the energy releases on the reactivity insertion rate, the Doppler reactivity feedback, the power form factor and the core size have been studied. The model enables a quick estimation of conservative values of the destructive mechanical energy releases following a nuclear explosion and gives suggestions as to how to reduce or even avoid such excursions

Chernobyl: recovery operations and the entombment of Reactor 4

International Nuclear Information System (INIS)

Dalziel, S.P.C.

1988-01-01

The immediate actions taken following the accident at the Chernobyl-number 4 reactor in April 1986 are described. These included actions to put out the fires, initial medical aid and the dropping of sand, lead, dolomite and boron onto the reactor from helicopters. Following this the chamber below the damaged reactor core was filled with concrete to prevent any further explosions or meltdown. The reactor was subsequently entombed in steel and concrete. The evacuation of the surrounding area is also mentioned. (U.K.)

Core fusion accidents in nuclear power reactors. Knowledge review

International Nuclear Information System (INIS)

Bentaib, Ahmed; Bonneville, Herve; Clement, Bernard; Cranga, Michel; Fichot, Florian; Koundy, Vincent; Meignen, Renaud; Corenwinder, Francois; Leteinturier, Denis; Monroig, Frederique; Nahas, Georges; Pichereau, Frederique; Van-Dorsselaere, Jean-Pierre; Cenerino, Gerard; Jacquemain, Didier; Raimond, Emmanuel; Ducros, Gerard; Journeau, Christophe; Magallon, Daniel; Seiler, Jean-Marie; Tourniaire, Bruno

2013-01-01

This reference document proposes a large and detailed review of severe core fusion accidents occurring in nuclear power reactors. It aims at presenting the scientific aspects of these accidents, a review of knowledge and research perspectives on this issue. After having recalled design and operation principles and safety principles for reactors operating in France, and the main studied and envisaged accident scenarios for the management of severe accidents in French PWRs, the authors describe the physical phenomena occurring during a core fusion accident, in the reactor vessel and in the containment building, their sequence and means to mitigate their effects: development of the accident within the reactor vessel, phenomena able to result in an early failure of the containment building, phenomena able to result in a delayed failure with the corium-concrete interaction, corium retention and cooling in and out of the vessel, release of fission products. They address the behaviour of containment buildings during such an accident (sizing situations, mechanical behaviour, bypasses). They review and discuss lessons learned from accidents (Three Mile Island and Chernobyl) and simulation tests (Phebus-PF). A last chapter gives an overview of software and approaches for the numerical simulation of a core fusion accident

Economic assessment of energetic scenarios

International Nuclear Information System (INIS)

Grandjean, Alain; Bureau, Dominique; Schubert, Katheline; Henriet, Fanny; Maggiar, Nicolas; Criqui, Patrick; Le Teno, Helene; Baumstark, Luc; Crassous, Renaud; Roques, Fabien

2013-09-01

This publication gathers contributions proposed by different members of the Economic Council for a Sustainable Development (CEDD) on the issue of energy transition, and more precisely on scenarios elaborated with respect to energy transition. A first set of contributions addresses models of energy transition (assessment of scenario costs to reach a factor 4; the issue of de-carbonation of energy consumption; study of ELECsim, a tool to highlight costs of scenarios of evolution of the electric power system). The second part addresses arbitrations and choice assessment (the importance of social and economic impacts of scenarios; challenges related to the joint definition of the discount rate and of the evolution of carbon value in time; the issue of assessment of the integration of renewable energies into the power system)

Nuclear Security Futures Scenarios

International Nuclear Information System (INIS)

Keller, Elizabeth James Kistin; Warren, Drake Edward; Hayden, Nancy Kay; Passell, Howard D.; Malczynski, Leonard A.; Backus, George A.

2017-01-01

This report provides an overview of the scenarios used in strategic futures workshops conducted at Sandia on September 21 and 29, 2016. The workshops, designed and facilitated by analysts in Center 100, used scenarios to enable thought leaders to think collectively about the changing aspects of global nuclear security and the potential implications for the US Government and Sandia National Laboratories.

Nuclear Security Futures Scenarios.

Energy Technology Data Exchange (ETDEWEB)

Keller, Elizabeth James Kistin [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Warren, Drake Edward [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hayden, Nancy Kay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Passell, Howard D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Malczynski, Leonard A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Backus, George A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-01-01

This report provides an overview of the scenarios used in strategic futures workshops conducted at Sandia on September 21 and 29, 2016. The workshops, designed and facilitated by analysts in Center 100, used scenarios to enable thought leaders to think collectively about the changing aspects of global nuclear security and the potential implications for the US Government and Sandia National Laboratories.

Core Professionalism Education in Surgery: A Systematic Review

Directory of Open Access Journals (Sweden)

Akile Sarıoğlu Büke

2018-03-01

Full Text Available Background: Professionalism education is one of the major elements of surgical residency education. Aims: To evaluate the studies on core professionalism education programs in surgical professionalism education. Study Design: Systematic review. Methods: This systematic literature review was performed to analyze core professionalism programs for surgical residency education published in English with at least three of the following features: program developmental model/instructional design method, aims and competencies, methods of teaching, methods of assessment, and program evaluation model or method. A total of 27083 articles were retrieved using EBSCOHOST, PubMed, Science Direct, Web of Science, and manual search. Results: Eight articles met the selection criteria. The instructional design method was presented in only one article, which described the Analysis, Design, Development, Implementation, and Evaluation model. Six articles were based on the Accreditation Council for Graduate Medical Education criterion, although there was significant variability in content. The most common teaching method was role modeling with scenario- and case-based learning. A wide range of assessment methods for evaluating professionalism education were reported. The Kirkpatrick model was reported in one article as a method for program evaluation. Conclusion: It is suggested that for a core surgical professionalism education program, developmental/instructional design model, aims and competencies, content, teaching methods, assessment methods, and program evaluation methods/models should be well defined, and the content should be comparable.

Core Professionalism Education in Surgery: A Systematic Review.

Science.gov (United States)

Sarıoğlu Büke, Akile; Karabilgin Öztürkçü, Özlem Sürel; Yılmaz, Yusuf; Sayek, İskender

2018-03-15

Professionalism education is one of the major elements of surgical residency education. To evaluate the studies on core professionalism education programs in surgical professionalism education. Systematic review. This systematic literature review was performed to analyze core professionalism programs for surgical residency education published in English with at least three of the following features: program developmental model/instructional design method, aims and competencies, methods of teaching, methods of assessment, and program evaluation model or method. A total of 27083 articles were retrieved using EBSCOHOST, PubMed, Science Direct, Web of Science, and manual search. Eight articles met the selection criteria. The instructional design method was presented in only one article, which described the Analysis, Design, Development, Implementation, and Evaluation model. Six articles were based on the Accreditation Council for Graduate Medical Education criterion, although there was significant variability in content. The most common teaching method was role modeling with scenario- and case-based learning. A wide range of assessment methods for evaluating professionalism education were reported. The Kirkpatrick model was reported in one article as a method for program evaluation. It is suggested that for a core surgical professionalism education program, developmental/instructional design model, aims and competencies, content, teaching methods, assessment methods, and program evaluation methods/models should be well defined, and the content should be comparable.

Scenario Methodology for Modelling of Future Landscape Developments as Basis for Assessing Ecosystem Services

Directory of Open Access Journals (Sweden)

Matthias Rosenberg

2014-04-01

Full Text Available The ecosystems of our intensively used European landscapes produce a variety of natural goods and services for the benefit of humankind, and secure the basics and quality of life. Because these ecosystems are still undergoing fundamental changes, the interest of the society is to know more about future developments and their ecological impacts. To describe and analyze these changes, scenarios can be developed and an assessment of the ecological changes can be carried out subsequently. In the project „Landscape Saxony 2050“; a methodology for the construction of exploratory scenarios was worked out. The presented methodology provides a possibility to identify the driving forces (socio-cultural, economic and ecological conditions of the landscape development. It allows to indicate possible future paths which lead to a change of structures and processes in the landscape and can influence the capability to provide ecosystem services. One essential component of the applied technique is that an approach for the assessment of the effects of the landscape changes on ecosystem services is integrated into the developed scenario methodology. Another is, that the methodology is strong designed as participatory, i.e. stakeholders are integrated actively. The method is a seven phase model which provides the option for the integration of the stakeholders‘ participation at all levels of scenario development. The scenario framework was applied to the district of Görlitz, an area of 2100 sq km located at the eastern border of Germany. The region is affected by strong demographic as well as economic changes. The core issue focused on the examination of landscape change in terms of biodiversity. Together with stakeholders, a trend scenario and two alternative scenarios were developed. The changes of the landscape structure are represented in story lines, maps and tables. On basis of the driving forces of the issue areas „cultural / social values“ and �

Core Hunter 3: flexible core subset selection.

Science.gov (United States)

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

2018-05-31

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

Assessment of nuclear power scenarios allowing for matrix behavior in radiological impact modeling of disposal scenarios

International Nuclear Information System (INIS)

Tronche, E.; Boussier, H.

2000-01-01

Under the provisions of the 1991 French radioactive waste management law, various fuel cycle scenarios will be assessed and compared in terms of feasibility, flexibility, cost, and ultimate waste radio-toxic inventory. The latter criterion may be further broken down into 'potential radio-toxic inventory' (the radio-toxic inventory of all the radionuclides produced) and 'residual radio-toxic inventory' (the radionuclide fraction reaching the biosphere after migration from the repository). The innovative scientific contribution of this study is to consider a third type of radio-toxic inventory: the potential radio-toxic inventory after conditioning, i.e. taking into account the containment capacity of the radionuclide conditioning matrices. The matrix fraction subjected to alteration over time determines the potential for radionuclide release, hence the notion of the potential radio-toxic inventory after conditioning. An initial comparison of possible scenarios is proposed by considering orders of magnitude for the radionuclide containment capacity of the disposal matrices and for their mobilization potential. All the scenarios investigated are normalized to the same annual electric power production so that a legitimate comparison can be established for the ultimate wasteform produced per year of operation. This approach reveals significant differences among the scenarios considered that do not appear when only the raw potential radio-toxic inventory is taken into account. The matrix containment performance has a decisive effect on the final impact of a given scenario or type of scenario. Pu recycling scenarios thus reduce the potential radio-toxicity by roughly a factor of 50 compared with an open cycle; the gain rises to a factor of about 300 for scenarios in which Pu and the minor actinides are recycled. Interestingly, the results obtained by the use of a dedicated containment matrix for the minor actinides in a scenario limited to Pu recycling were comparable to

Improved core electron confinement on JET

International Nuclear Information System (INIS)

Litaudon, X.; Baranov, Y.; Voitsekhovitch, I.

1999-01-01

Formation of core regions with reduced electron transport is reported in regimes with current profile shaping at JET. The electron heat diffusivity (Χ c ) is reduced down to 0.5 m 2 /s in the region of low magnetic shear with an ICRH power of 1 MW with no indication of a threshold. In the high performance optimised shear regime, obtained in scenarios dominated by ion heating, internal transport barriers on the ion temperature profiles are simultaneously accompanied by a significant reduction of the electron heat diffusivity at two-third of the plasma radius. In this regime, recent results and measurements obtained with the new gas-box divertor configuration are reported together with their transport analyses. The results indicate that Χ c is reduced by one order of magnitude in a spatially localised region. (authors)

Inner Core Tilt and Polar Motion: Probing the Dynamics Deep Inside the Earth

Science.gov (United States)

Dumberry, M.; Bloxham, J.

2003-12-01

A tilted inner core permits exchange of angular momentum between the core and the mantle through gravitational and pressure torques and, as a result, changes in the direction of Earth's axis of rotation with respect to the mantle. Some of the observed variations in the direction of Earth's rotation could then be caused by equatorial torques on the inner core which tilt the latter out of its alignment with the mantle. In this work, we investigate whether such a scenario could explain the decade polar motion known as the Markowitz wobble. We show that a decade polar motion of the same amplitude as the observed Markowitz wobble requires a torque of 1020 N m which tilts the inner core by 0.07 degrees. This result critically depends on the viscosity of the inner core; for a viscosity less than 5 x 1017 Pa s, larger torques are required. A torque of 1020 N m with decadal periodicity can perhaps be produced by electromagnetic coupling between the inner core and a component of the flow in the outer core known as torsional oscillations, provided that the radial magnetic field at the inner core boundary is on the order of 3 to 4 mT and satisfies certain geometrical constraints. The resulting polar motion thus produced is eccentric and polarized, in agreement with the observations. Our model suggests that equatorial torques at the inner core boundary might also excite the Chandler wobble, provided shorter wavelength torsional oscillations with higher natural frequencies have enough power or provided there exists another physical mechanism that can generate a large torque at a 14 month period.

The structure of protostellar dense cores: a millimeter continuum study

International Nuclear Information System (INIS)

Motte, Frederique

1998-01-01

A comprehensive theoretical scenario explains low-mass star formation and describes the gravitational collapse of an isolated 'ideal' dense core. The major aim of this thesis is to check the standard model predictions on the structure of protostellar dense cores (or envelopes). The earliest stages of star formation remain poorly known because the protostars are still deeply embedded in massive, opaque circumstellar cocoons. On the one hand, sensitive bolometer arrays very recently allow us to measure the millimeter continuum emission arising from dense cores. Such observations are a powerful tool to constrain the density structure of proto-stellar dense cores (on large length scale). In particular, we studied the structure of isolated proto-stellar envelopes in Taurus and protostars in the ρ Ophiuchi cluster. In order to accurately derive their envelope density power law, we simulated the observation of several envelope models. Then we show that most of the Taurus protostars present a density structure consistent with the standard model predictions. In contrast, dense cores in ρ Ophiuchi main cloud are highly fragmented and protostellar envelope have finite size. Moreover fragmentation appears to be essential in determining the final stellar mass of ρ Oph forming stars. In clusters, fragmentation may thus be at the origin of the stellar initial mass function (IMF). On the other hand, our interferometric millimeter continuum observations are tracing (with higher angular resolution) the inner part of protostellar envelopes. Our study show that disks during protostellar stages are not yet massive and thus do not perturb the analysis of envelope density structure. (author) [fr

Climate scenarios for California

Science.gov (United States)

Cayan, Daniel R.; Maurer, Ed; Dettinger, Mike; Tyree, Mary; Hayhoe, Katharine; Bonfils, Celine; Duffy, Phil; Santer, Ben

2006-01-01

Possible future climate changes in California are investigated from a varied set of climate change model simulations. These simulations, conducted by three state-of-the-art global climate models, provide trajectories from three greenhouse gas (GHG) emission scenarios. These scenarios and the resulting climate simulations are not “predictions,” but rather are a limited sample from among the many plausible pathways that may affect California’s climate. Future GHG concentrations are uncertain because they depend on future social, political, and technological pathways, and thus the IPCC has produced four “families” of emission scenarios. To explore some of these uncertainties, emissions scenarios A2 (a medium-high emissions) and B1 (low emissions) were selected from the current IPCC Fourth climate assessment, which provides several recent model simulations driven by A2 and B1 emissions. The global climate model simulations addressed here were from PCM1, the Parallel Climate Model from the National Center for Atmospheric Research (NCAR) and U.S. Department of Energy (DOE) group, and CM2.1 from the National Oceanic and Atmospheric Administration (NOAA) Geophysical Fluids Dynamics Laboratory (GFDL).

Development of internal transport barrier scenarios at ITER-relevant high triangularity in JET

International Nuclear Information System (INIS)

Rimini, F.G.; Becoulet, M.; Giovannozzi, E.; Lomas, P.J.; Tudisco, O.; Alper, B.; Crisanti, F.; Baar, M. de; Luna, E. de La; Vries, P. de; Ekedahl, A.; Hawkes, N.; Huysmans, G.; Litaudon, X.; Parail, V.; Saibene, G.; Tuccillo, A.A.; Zastrow, K.D.

2005-01-01

The development of scenarios characterized by H-mode confinement and internal transport barriers (ITBs) in high triangularity, δ ∼ 0.4-0.5, discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours H-modes which are ELM-free or develop type I edge localized mode (ELM) activity, which inhibits long lasting ITBs. The recent experiments reported here concentrate on integrated optimization of edge and core conditions. The stability of the edge pedestal was controlled using gas injection, deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, resulting in L-mode. In parallel, the conditions for triggering and sustaining ITBs encompassing a large proportion of the plasma volume (outer ITBs) were optimized, as opposed to less performing ITBs located closer to the plasma centre (inner ITB). These plasmas have deeply reversed target current profiles with q min ∼ 3 and a narrow inner ITB, located typically at a small normalized radius ρ E , at q 95 = 7.5, H 89 β N ∼ 3.5-4 and ∼60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios

Scenario development methods and practice

International Nuclear Information System (INIS)

2001-01-01

The safe management of radioactive waste is an essential aspect of all nuclear power programmes. Although a general consensus has been reached in OECD countries on the use of geological repositories for the disposal of high-level radioactive waste, analysis of the long-term safety of these repositories, using performance assessment and other tools, is required prior to implementation. The initial stage in developing a repository safety assessment is the identification of all factors that may be relevant to the long-term safety of the repository and their combination to form scenarios. This must be done in a systematic and transparent way in order to assure the regulatory authorities that nothing important has been forgotten. Scenario development has become the general term used to describe the collection and organisation of the scientific and technical information necessary to assess the long-term performance or safety of radioactive waste disposal systems. This includes the identification of the relevant features, events and processes (FEPs), the synthesis of broad models of scientific understanding, and the selection of cases to be calculated. Scenario development provides the overall framework in which the cases and their calculated consequences can be discussed, including biases or shortcomings due to omissions or lack of knowledge. The NEA Workshop on Scenario Development was organised in Madrid, in May 1999, with the objective of reviewing developments in scenario methodologies and applications in safety assessments since 1992. The outcome of this workshop is the subject of this book. It is a review of developments in scenario methodologies based on a large body of practical experience in safety assessments. It will be of interest to radioactive waste management experts as well as to other specialists involved in the development of scenario methodologies. (author)

Experiments on determination and limitation of fission and activation product release during core meltdown

Energy Technology Data Exchange (ETDEWEB)

Albrecht, H; Krause, W; Wild, H [Kernforschungszentrum Karlsruhe (Germany, F.R.). Inst. fuer Radiochemie; Perinic, D; Kammerer, B; Knauss, H; Mack, A; Stuka, B [Kernforschungszentrum Karlsruhe (Germany, F.R.). Abt. Reaktorbetrieb und Technik; Osborne, M F

1976-06-01

In melting experiments with Corium samples of 500 g under air, temperatures of about 2,300/sup 0/C could be reached whereas 500 g samples of Ni could be heated in argon up to only 1,400/sup 0/C. Obviously, the exothermic oxidation reaction of the zircaloy and steel is the reason for that considerable rise of the Corium temperature in air. Using smaller Ni samples (30 g) the maximum HF-power being coupled to the melt material has been determined by measuring the time of constant temperature at the melting point with the generator at full power. The thermal power effectively transmitted to the specimen was only 1.2 KW. Thus, for melting kg-amounts of Corium, a higher generator power is needed as well as a more effective HF-coupling. The rented generator now in use will be replaced, therefore, by a device with a nominal power of 120 KW, and the power transmission to the induction coil will be improved. The first series of release experiments with 30-60 g of Corium will not be affected by these changes. Melting experiments with various steel components (Fe, Cr, Ni) in an argon atmosphere and with Corium under air did not show any selective influence of the vapors on the intensities of the wavelengths used for the temperature measurement. The behavior of a spherical glass vessel above the crucible was quite satisfactory. After melting tests with temperatures of more than 2,700/sup 0/C it was not broken nor did it show any appearance of softening.

Safety device for nuclear fission reactors

International Nuclear Information System (INIS)

Brownlee, M.L.

1982-01-01

A plurality of radially arranged and neutron absorbing baffles are stacked in vertical sets under the fuel core assemblies, and the whole enclosed in a bottle shaped containment vessel. The radially arranged baffles of each set extend vertically, and each set has double the number of baffles as the set above it in the stack. A melt-down of a fuel core assembly drops the fissioning nuclear fuel into the stacked sets of baffles, there, as it passes through, to be progressively divided, redivided and dispersed in smaller and smaller masses between the doubling number of baffles in safe fuel pellet size. Neutron absorbing containment prevents contamination of the environment and together with cooling means stops fissioning of fuel

Asset Meltdown

DEFF Research Database (Denmark)

Marekwica, Marcel; Maurer, Raimond; Sebastian, Steffen P.

2011-01-01

Executive Summary. This paper analyzes the relation between demographic structure and real asset returns on Treasury bills, bonds, and stocks for the G7 countries (United States, Canada, Japan, Italy, France, the United Kingdom, and Germany). A macroeconomic multifactor model is used to examine a...

Multiscale scenarios for nature futures

NARCIS (Netherlands)

Rosa, Isabel M.D.; Pereira, Henrique M.; Ferrier, Simon; Alkemade, Rob; Acosta, Lilibeth A.; Akcakaya, H. Resit; Den Belder, Eefje; Fazel, Asghar M.; Fujimori, Shinichiro; Harfoot, Mike; Harhash, Khaled A.; Harrison, Paula A.; Hauck, Jennifer; Hendriks, Rob J.J.; Hernández, Gladys; Jetz, Walter; Karlsson-Vinkhuyzen, Sylvia I.; Kim, Hyejin; King, Nicholas; Kok, Marcel T.J.; Kolomytsev, Grygoriy O.; Lazarova, Tanya; Leadley, Paul; Lundquist, Carolyn J.; García Márquez, Jaime; Meyer, Carsten; Navarro, Laetitia M.; Nesshöver, Carsten; Ngo, Hien T.; Ninan, Karachepone N.; Palomo, Maria G.; Pereira, Laura M.; Peterson, Garry D.; Pichs, Ramon; Popp, Alexander; Purvis, Andy; Ravera, Federica; Rondinini, Carlo; Sathyapalan, Jyothis; Schipper, Aafke M.; Seppelt, Ralf; Settele, Josef; Sitas, Nadia; Van Vuuren, Detlef|info:eu-repo/dai/nl/11522016X

2017-01-01

Targets for human development are increasingly connected with targets for nature, however, existing scenarios do not explicitly address this relationship. Here, we outline a strategy to generate scenarios centred on our relationship with nature to inform decision-making at multiple scales.

Trust and the illusive force of scenarios

DEFF Research Database (Denmark)

Selin, Cynthia Lea

2006-01-01

formulation and decision-making. By definition, scenarios are possible versions of the future so judging and evaluating scenarios is thus not about revealing truthfulness, but rather demonstrating trust, reliability, credibility in the absence of truth and in the face of varied influences and possible...... becomes interesting is how scenarios convey authority and trustworthiness. How is it that scenarios attain and maintain power to compel people to action, change their worldview, or influence the directions of decisions or consensus? This piece examines the process, participation and products of scenario...

Comparison of traditional advanced cardiac life support (ACLS) course instruction vs. a scenario-based, performance oriented team instruction (SPOTI) method for Korean paramedic students.

Science.gov (United States)

Lee, Christopher C; Im, Mark; Kim, Tae Min; Stapleton, Edward R; Kim, Kyuseok; Suh, Gil Joon; Singer, Adam J; Henry, Mark C

2010-01-01

Current Advanced Cardiac Life Support (ACLS) course instruction involves a 2-day course with traditional lectures and limited team interaction. We wish to explore the advantages of a scenario-based performance-oriented team instruction (SPOTI) method to implement core ACLS skills for non-English-speaking international paramedic students. The objective of this study was to determine if scenario-based, performance-oriented team instruction (SPOTI) improves educational outcomes for the ACLS instruction of Korean paramedic students. Thirty Korean paramedic students were randomly selected into two groups. One group of 15 students was taught the traditional ACLS course. The other 15 students were instructed using a SPOTI method. Each group was tested using ACLS megacode examinations endorsed by the American Heart Association. All 30 students passed the ACLS megacode examination. In the traditional ACLS study group an average of 85% of the core skills were met. In the SPOTI study group an average of 93% of the core skills were met. In particular, the SPOTI study group excelled at physical examination skills such as airway opening, assessment of breathing, signs of circulation, and compression rates. In addition, the SPOTI group performed with higher marks on rhythm recognition compared to the traditional group. The traditional group performed with higher marks at providing proper drug dosages compared to the SPOTI students. However, the students enrolled in the SPOTI method resulted in higher megacode core compliance scores compared to students trained in traditional ACLS course instruction. These differences did not achieve statistical significance due to the small sample size. Copyright 2010 Elsevier Inc. All rights reserved.

The role of fusion power in energy scenarios. Proposed method and review of existing scenarios

International Nuclear Information System (INIS)

Lako, P; Ybema, J.R.; Seebregts, A.J.

1998-04-01

The European Commission wishes more insight in the potential role of fusion energy in the second half of the 21st century. Therefore, several scenario studies are carried out in the so-called macro-task Long Term Scenarios to investigate the potential role of fusion power in the energy system. The main contribution of ECN to the macro-task is to perform a long term energy scenario study for Western Europe with special focus on the role of fusion power. This interim report gives some methodological considerations for such an analysis. A discussion is given on the problems related to the long time horizon of the scenario study such as the forecast of technological innovations, the selection of appropriate discount rates and the links with climate change. Key parameters which are expected to have large effects on the role and cost-effectiveness are discussed in general terms. The key parameters to be varied include level and structure of energy demand, availability and prices of fossil energy, CO2 reduction policy, discount rates, cost and potential of renewable energy sources, availability of fission power and CO2 capture and disposal and the cost and the maximum rate of market growth of fusion power. The scenario calculations are to be performed later in the project with the help of an existing cost minimisation model of the Western European energy system. This MARKAL model is briefly introduced. The results of the model calculations are expected to make clear under which combinations of scenario parameters fusion power is needed and how large the expected financial benefits will be. The present interim report also gives an evaluation of existing energy scenarios with respect to the role of fusion power. 18 refs

Analysis Of Core Management For The Transition Cores Of RSG-GAS Reactor To Full-Silicide Core

International Nuclear Information System (INIS)

Malem Sembiring, Tagor; Suparlina, Lily; Tukiran

2001-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 g U/cc is still doing. At the end of 2000, the reactor has been operated for 3 transition cores which is the mixed core of oxide-silicide. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 10 transition cores to achieve a full-silicide core. The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters such as excess reactivity and shutdown margin. The measurement of the core parameters was carried out using the method of compensation of couple control rods. The experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safety to a full-silicide core

Multiscale scenarios for nature futures

NARCIS (Netherlands)

Rosa, Isabel M.D.; Pereira, Henrique Miguel; Ferrier, Simon; Alkemade, J.R.M.; Acosta, Lilibeth A.; Resit Akcakaya, H.; Belder, den E.; Fazel, Asghar M.; Fujimori, Shinichiro; Harfoot, Mike; Harhash, Khaled A.; Harrison, Paula A.; Hauck, Jennifer; Hendriks, Rob J.J.; Hernández, Gladys; Jetz, Walter; Karlsson-Vinkhuyzen, S.I.S.E.; Kim, Hyejin; King, Nicholas; Kok, Marcel; Kolomytsev, Grygoriy O.; Lazarova, Tanya; Leadley, Paul; Lundquist, Carolyn J.; García Márquez, Jaime; Meyer, Carsten; Navarro, Laetitia M.; Nesshöver, Carsten; Ngo, Hien T.; Ninan, Karachepone N.; Palomo, Maria G.; Pereira, Laura; Peterson, G.D.; Pichs, Ramon; Popp, Alexander; Purvis, Andy; Ravera, Federica; Rondinini, Carlo; Sathyapalan, Jyothis; Schipper, Aafke; Seppelt, Ralf; Settele, Josef; Sitas, Nadia; Vuuren, van D.

2017-01-01

Targets for human development are increasingly connected with targets for nature, however, existing scenarios do not explicitly address this relationship. Here, we outline a strategy to generate scenarios centred on our relationship
with nature to inform decision-making at multiple scales.

Core-to-core uniformity improvement in multi-core fiber Bragg gratings

Science.gov (United States)

Lindley, Emma; Min, Seong-Sik; Leon-Saval, Sergio; Cvetojevic, Nick; Jovanovic, Nemanja; Bland-Hawthorn, Joss; Lawrence, Jon; Gris-Sanchez, Itandehui; Birks, Tim; Haynes, Roger; Haynes, Dionne

2014-07-01

Multi-core fiber Bragg gratings (MCFBGs) will be a valuable tool not only in communications but also various astronomical, sensing and industry applications. In this paper we address some of the technical challenges of fabricating effective multi-core gratings by simulating improvements to the writing method. These methods allow a system designed for inscribing single-core fibers to cope with MCFBG fabrication with only minor, passive changes to the writing process. Using a capillary tube that was polished on one side, the field entering the fiber was flattened which improved the coverage and uniformity of all cores.

Interfacing high-fidelity core neutronics models to whole plant models

International Nuclear Information System (INIS)

McEllin, M.

1999-01-01

Until recently available computer power dictated that whole-plant models of nuclear power stations have typically employed simple models of the reactor core which can not match the fidelity of safety-qualified 2-group, 3D neutronics models. As a result the treatment of situations involving strong coupling between the core and the rest of the plant has inevitably been somewhat approximate, requiring conservative modelling assumptions, or manual iteration between cases, to bound worse case scenarios. Such techniques not only place heavy demands on the engineers involved, they may also result in potentially unnecessary operational constraints. Hardware is today no longer the limiting factor, but the cost of developing and validating high-quality software is now such that it appears attractive to build new systems with a wider simulation scope by using existing stand-alone codes as sub-components. This is not always as straightforward as it might at first appear. This paper illustrates some of the pitfalls, and discusses more sophisticated and robust strategies. (author)

TURVA-2012: Formulation of radionuclide release scenarios

International Nuclear Information System (INIS)

Marcos, Nuria; Hjerpe, Thomas; Snellman, Margit; Ikonen, Ari; Smith, Paul

2014-01-01

TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR) and application for a construction licence for a repository for disposal of spent nuclear fuel at the Olkiluoto site in south-western Finland. This paper gives a summary of the scenarios and the methodology followed in formulating them as described in TURVA-2012: Formulation of Radionuclide Release Scenarios (Posiva, 2013). The scenarios are further analysed in TURVA-2012: Assessment of Radionuclide Release Scenarios for the Repository System and TURVA-2012: Biosphere Assessment (Posiva, 2012a, 2012b). The formulation of scenarios takes into account the safety functions of the main barriers of the repository system and the uncertainties in the features, events, and processes (FEP) that may affect the entire disposal system (i.e. repository system plus the surface environment) from the emplacement of the first canister until the far future. In the report TURVA-2012: Performance Assessment (2012d), the performance of the engineered and natural barriers has been assessed against the loads expected during the evolution of the repository system and the site. Uncertainties have been identified and these are taken into account in the formulation of radionuclide release scenarios. The uncertainties in the FEP and evolution of the surface environment are taken into account in formulating the surface environment scenarios used ultimately in estimating radiation exposure. Formulating radionuclide release scenarios for the repository system links the reports Performance Assessment and Assessment of Radionuclide Release Scenarios for the Repository System. The formulation of radionuclide release scenarios for the surface environment brings together biosphere description and the surface environment FEP and is the link to the assessment of the surface environment scenarios summarised in TURVA-2012: Biosphere Assessment. (authors)

Automated Analysis of Infinite Scenarios

DEFF Research Database (Denmark)

Buchholtz, Mikael

2005-01-01

The security of a network protocol crucially relies on the scenario in which the protocol is deployed. This paper describes syntactic constructs for modelling network scenarios and presents an automated analysis tool, which can guarantee that security properties hold in all of the (infinitely many...

Accident scenario diagnostics with neural networks

International Nuclear Information System (INIS)

Guo, Z.

1992-01-01

Nuclear power plants are very complex systems. The diagnoses of transients or accident conditions is very difficult because a large amount of information, which is often noisy, or intermittent, or even incomplete, need to be processed in real time. To demonstrate their potential application to nuclear power plants, neural networks axe used to monitor the accident scenarios simulated by the training simulator of TVA's Watts Bar Nuclear Power Plant. A self-organization network is used to compress original data to reduce the total number of training patterns. Different accident scenarios are closely related to different key parameters which distinguish one accident scenario from another. Therefore, the accident scenarios can be monitored by a set of small size neural networks, called modular networks, each one of which monitors only one assigned accident scenario, to obtain fast training and recall. Sensitivity analysis is applied to select proper input variables for modular networks

Scenarios for Ecosystem Services: An Overview

Directory of Open Access Journals (Sweden)

Stephen R. Carpenter

2006-06-01

Full Text Available The Millennium Ecosystem Assessment (MA scenarios address changes in ecosystem services and their implications for human well-being. Ecological changes pose special challenges for long-term thinking, because of the possibility of regime shifts that occur rapidly yet alter the availability of ecosystem services for generations. Moreover, ecological feedbacks can intensify human modification of ecosystems, creating a spiral of poverty and ecosystem degradation. Such complex dynamics were evaluated by a mixture of qualitative and quantitative analyses in the MA scenarios. Collectively, the scenarios explore problems such as the connections of poverty reduction and ecosystem services, and trade-offs among ecosystem services. Several promising approaches are considered by the scenarios, including uses of biodiversity to build resilience of ecosystem services, actively adaptive management, and green technology. Although the scenarios do not prescribe an optimal path, they illuminate the consequences of different policies toward ecosystem services.

Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

International Nuclear Information System (INIS)

Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

2007-01-01

Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

Plant incident progress scenario analysis for nuclear emergency exercise of Fukui Pref. 2009

International Nuclear Information System (INIS)

Tatebe, Yasumasa; Yoshida, Yoshitaka; Minami, Noritoshi

2010-01-01

Severe accident scenario was analyzed to assist the nuclear emergency exercise of Fukui prefecture in 2009. The evaluation of the postulated disaster was evaluated based on analysis result, including the time of occurrence of specified events, response of the plant to the AM measures, etc. The evaluation results are as follows: (1) loss of core cooling took place about 7 hours after, abnormal leakage of fission products from containment vessel to the containment annulus started about 17 hours after and the disaster terminated about 46 hours after the initiation of the accident, (2) although multiple failures of the engineered safety system and failure of AM measures were postulated in the accident scenario, the integrity of the containment vessel could be maintained successfully for a long term, if the AM measures are taken to protect the containment vessel. (3) according to the evaluation of radiation dose rate during the emergency response assuming the operation of recirculation mode using containment spray A, the dose rate could be too high for the operators to work not only in the containment spray A area but also in the another containment spray heat exchanger and the residual heat removal system heat exchanger areas. (author)

MORECA: A computer code for simulating modular high-temperature gas-cooled reactor core heatup accidents

International Nuclear Information System (INIS)

Ball, S.J.

1991-10-01

The design features of the modular high-temperature gas-cooled reactor (MHTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. This report describes the ORNL MORECA code, which was developed for analyzing postulated long-term core heatup scenarios for which active cooling systems used to remove afterheat following the accidents can be assumed to the unavailable. Simulations of long-term loss-of-forced-convection accidents, both with and without depressurization of the primary coolant, have shown that maximum core temperatures stay below the point at which any significant fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. MORECA models the US Department of Energy reference design of a standard MHTGR

Exchange of transverse plasmons and electrical conductivity of neutron star cores

International Nuclear Information System (INIS)

Shternin, P. S.

2008-01-01

We study the electrical conductivity in magnetized neutron star cores produced by collisions between charged particles. We take into account the ordinary exchange of longitudinal plasmons and the exchange of transverse plasmons in collisions between particles. The exchange of transverse plasmons is important for collisions between relativistic particles, but it has been disregarded previously when calculating the electrical conductivity. We show that taking this exchange into account changes the electrical conductivity, including its temperature dependence (thus, for example, the temperature dependence of the electrical resistivity along the magnetic field in the low-temperature limit takes the form R parallel ∝ T 5/3 instead of the standard dependence R parallel ∝ T 2 for degenerate Fermi systems). We briefly describe the effect of possible neutron and proton superfluidity in neutron star cores on the electrical conductivity and discuss various scenarios for the evolution of neutron star magnetic fields

Coupling between core and cladding modes in a helical core fiber with large core offset

International Nuclear Information System (INIS)

Napiorkowski, Maciej; Urbanczyk, Waclaw

2016-01-01

We analyzed the effect of resonant coupling between core and cladding modes in a helical core fiber with large core offset using the fully vectorial method based on the transformation optics formalism. Our study revealed that the resonant couplings to lower order cladding modes predicted by perturbative methods and observed experimentally in fibers with small core offsets are in fact prohibited for larger core offsets. This effect is related to the lack of phase matching caused by elongation of the optical path of the fundamental modes in the helical core. Moreover, strong couplings to the cladding modes of the azimuthal modal number much higher than predicted by perturbative methods may be observed for large core offsets, as the core offset introduces higher order angular harmonics in the field distribution of the fundamental modes. Finally, in contrast to previous studies, we demonstrate the existence of spectrally broad polarization sensitive couplings to the cladding modes suggesting that helical core fibers with large core offsets may be used as broadband circular polarizers. (paper)

Room temperature nanojoining of Cu-Ag core-shell nanoparticles and nanowires

Energy Technology Data Exchange (ETDEWEB)

Wang, Jiaqi; Shin, Seungha, E-mail: sshin@utk.edu [The University of Tennessee, Department of Mechanical, Aerospace and Biomedical Engineering (United States)

2017-02-15

Room temperature (T{sub room}, 300 K) nanojoining of Ag has been widely employed in fabrication of microelectronic applications where the shapes and structures of microelectronic components must be maintained. In this research, the joining processes of pure Ag nanoparticles (NPs), Cu-Ag core-shell NPs, and nanowires (NWs) are studied using molecular dynamics simulations at T{sub room}. The evolution of densification, potential energy, and structural deformation during joining process are analyzed to identify joining mechanisms. Depending on geometry, different joining mechanisms including crystallization-amorphization, reorientation, Shockley partial dislocation are determined. A three-stage joining scenario is observed in both joining process of NPs and NWs. Besides, the Cu core does not participate in all joining processes, however, it enhances the mobility of Ag shell atoms, contributing to a higher densification and bonding strength at T{sub room}, compared with pure Ag nanomaterials. The tensile test shows that the nanojoint bears higher rupture strength than the core-shell NW itself. This study deepens understanding in the underlying joining mechanisms and thus nanojoint with desirable thermal, electrical, and mechanical properties could be potentially achieved.

Evaluation of In-Core Fuel Management for the Transition Cores of RSG-GAS Reactor to Full-Silicide Core

International Nuclear Information System (INIS)

S, Tukiran; MS, Tagor; P, Surian

2003-01-01

The core conversion of RSG-GAS reactor from oxide to silicide core with meat density of 2.96 gU/cc has been done. The core-of RSG-GAS reactor has been operated full core of silicide fuels which is started with the mixed core of oxide-silicide start from core 36. Based on previous work, the calculated core parameter for the cores were obtained and it is needed 9 transition cores (core 36 - 44) to achieve a full-silicide core (core 45). The objective of this work is to acquire the effect of the increment of the number of silicide fuel on the core parameters. Conversion core was achieved by transition cores mixed oxide-silicide fuels. Each transition core is calculated and measured core parameter such as, excess reactivity and shutdown margin. Calculation done by Batan-EQUIL-2D code and measurement of the core parameters was carried out using the method of compensation of couple control rods. The results of calculation and experiment shows that the excess reactivity trends lower with the increment of the number of silicide fuel in the core. However, the shutdown margin is not change with the increment of the number of silicide fuel. Therefore, the transition cores can be operated safely to a full-silicide core

Sulfur Saturation Limits in Silicate Melts and their Implications for Core Formation Scenarios for Terrestrial Planets

Science.gov (United States)

Holzheid, Astrid; Grove, Timothy L.

2002-01-01

This study explores the controls of temperature, pressure, and silicate melt composition on S solubility in silicate liquids. The solubility of S in FeO-containing silicate melts in equilibrium with metal sulfide increases significantly with increasing temperature but decreases with increasing pressure. The silicate melt structure also exercises a control on S solubility. Increasing the degree of polymerization of the silicate melt structure lowers the S solubility in the silicate liquid. The new set of experimental data is used to expand the model of Mavrogenes and O'Neill(1999) for S solubility in silicate liquids by incorporating the influence of the silicate melt structure. The expected S solubility in the ascending magma is calculated using the expanded model. Because the negative pressure dependence of S solubility is more influential than the positive temperature dependence, decompression and adiabatic ascent of a formerly S-saturated silicate magma will lead to S undersaturation. A primitive magma that is S-saturated in its source region will, therefore, become S-undersaturated as it ascends to shallower depth. In order to precipitate magmatic sulfides, the magma must first cool and undergo fractional crystallization to reach S saturation. The S content in a metallic liquid that is in equilibrium with a magma ocean that contains approx. 200 ppm S (i.e., Earth's bulk mantle S content) ranges from 5.5 to 12 wt% S. This range of S values encompasses the amount of S (9 to 12 wt%) that would be present in the outer core if S is the light element. Thus, the Earth's proto-mantle could be in equilibrium (in terms of the preserved S abundance) with a core-forming metallic phase.

Nuclear Safety Project. Annual report 1986

International Nuclear Information System (INIS)

1987-09-01

The annual report 1986 is a detailed description of work within the Nuclear Safety Project performed in 1986 in the nuclear safety field by KfK institutes and departments and by external institutes on behalf of KfK. It includes individual research activities on dynamic loads and strains of reactor components under accident conditions, fuel behaviour under accident conditions, investigation and control of LWR core-meltdown accidents, improvement of fission product retention and reduction of radiation exposure, and on behaviour, impact and removal of released pollutants. (DG)

Useful global-change scenarios: current issues and challenges

International Nuclear Information System (INIS)

Parson, E A

2008-01-01

Scenarios are increasingly used to inform global-change debates, but their connection to decisions has been weak and indirect. This reflects the greater number and variety of potential users and scenario needs, relative to other decision domains where scenario use is more established. Global-change scenario needs include common elements, e.g., model-generated projections of emissions and climate change, needed by many users but in different ways and with different assumptions. For these common elements, the limited ability to engage diverse global-change users in scenario development requires extreme transparency in communicating underlying reasoning and assumptions, including probability judgments. Other scenario needs are specific to users, requiring a decentralized network of scenario and assessment organizations to disseminate and interpret common elements and add elements requiring local context or expertise. Such an approach will make global-change scenarios more useful for decisions, but not less controversial. Despite predictable attacks, scenario-based reasoning is necessary for responsible global-change decisions because decision-relevant uncertainties cannot be specified scientifically. The purpose of scenarios is not to avoid speculation, but to make the required speculation more disciplined, more anchored in relevant scientific knowledge when available, and more transparent.

Input parameters and scenarios, including economic inputs

DEFF Research Database (Denmark)

Boklund, Anette; Hisham Beshara Halasa, Tariq

2012-01-01

scenarios, we excluded hobby-type farms1 In the vaccination scenarios, herds within the vaccination zone were simulated to be vaccinated 14 days after detection of the first herd or when 10, 20, 30 or 50 herds were infected. All herds within the zones were simulated to be vaccinated. We used vaccination...... zones of either a 1, 2, 3 or 5 km. In some scenarios, hobby herds were not vaccinated. In one scenario, no sheep were vaccinated, and in another scenario no swine were vaccinated. from depopulation in zones. The resources for depopulation were estimated to 4,800 swine and 2,000 ruminants a day...

FEATURES OF FORMATION SCENARIOS OF THE FOOD INDUSTRY

Directory of Open Access Journals (Sweden)

Serhiy Nozhenko

2016-03-01

Full Text Available The article is devoted to the topic – the invention of approaches to enterprise development oil and fat industry. The goal – to identify features of formation scenarios for the food industry. The chosen methodology allowed to analyze and evaluate the functioning of state enterprises. Results are discovering the benefits of scenario planning to find ways of business development. Novelty – a rationale for the use of scenario planning for enterprise development oil and fat industry, developing a set of scenarios. The practical significance – providing specific scenarios for enterprises of oil and fat industry. Keywords: scenario planning, scenario development, enterprise, oil and fat industry outlook. JEL: L 66

FAST Plasma Scenarios and Equilibrium Configurations

International Nuclear Information System (INIS)

Calabro, G.; Crisanti, F.; Ramogida, G.; Cardinali, A.; Cucchiaro, A.; Maddaluno, G.; Pizzuto, A.; Pericoli Ridolfini, V.; Tuccillo, A.A.; Zonca, F.; Albanese, R.; Granucci, G.; Nowak, S.

2008-01-01

In this paper we present the Fusion Advanced Studies Torus (FAST) plasma scenarios and equilibrium configurations, designed to reproduce the ITER ones (with scaled plasma current) and suitable to fulfil plasma conditions for integrated studies of burning plasma physics, Plasma Wall interaction, ITER relevant operation problems and Steady State scenarios. The attention is focused on FAST flexibility in terms of both performance and physics that can be investigated: operations are foreseen at a wide range of parameters from high performance H-Mode (toroidal field, B T , up to 8.5 T; plasma current, I P , up to 8 MA) to advanced tokamak (AT) operation (I P =3 MA) as well as full non inductive current scenario (I P =2 MA). The coupled heating power is provided with 30MW delivered by an Ion Cyclotron Resonance Heating (ICRH) system (30-90MHz), 6 MW by a Lower Hybrid (LH) system (3.7 or 5 GHz) for the long pulse AT scenario, 4 MW by an Electron Cyclotron Resonant Heating (ECRH) system (170 GHz-B T =6T) for MHD and electron heating localized control and, eventually, with 10 MW by a Negative Ion Beam (NNBI), which the ports are designed to accommodate. In the reference H-mode scenario FAST preserves (with respect to ITER) fast ions induced as well as turbulence fluctuation spectra, thus, addressing the cross-scale couplings issue of micro- to meso-scale physics. The noninductive scenario at I P =2MA is obtained with 60-70 % of bootstrap and the remaining by LHCD. Predictive simulations of the H-mode scenarios described above have been performed by means of JETTO code, using a semi-empirical mixed Bohm/gyro-Bohm transport model. Plasma position and Shape Control studies are also presented for the reference scenario

Towards asteroseismology of core-collapse supernovae with gravitational-wave observations - I. Cowling approximation

Science.gov (United States)

Torres-Forné, Alejandro; Cerdá-Durán, Pablo; Passamonti, Andrea; Font, José A.

2018-03-01

Gravitational waves from core-collapse supernovae are produced by the excitation of different oscillation modes in the protoneutron star (PNS) and its surroundings, including the shock. In this work we study the relationship between the post-bounce oscillation spectrum of the PNS-shock system and the characteristic frequencies observed in gravitational-wave signals from core-collapse simulations. This is a fundamental first step in order to develop a procedure to infer astrophysical parameters of the PNS formed in core-collapse supernovae. Our method combines information from the oscillation spectrum of the PNS, obtained through linear perturbation analysis in general relativity of a background physical system, with information from the gravitational-wave spectrum of the corresponding non-linear, core-collapse simulation. Using results from the simulation of the collapse of a 35 M⊙ pre-supernova progenitor we show that both types of spectra are indeed related and we are able to identify the modes of oscillation of the PNS, namely g-modes, p-modes, hybrid modes, and standing accretion shock instability (SASI) modes, obtaining a remarkably close correspondence with the time-frequency distribution of the gravitational-wave modes. The analysis presented in this paper provides a proof of concept that asteroseismology is indeed possible in the core-collapse scenario, and it may serve as a basis for future work on PNS parameter inference based on gravitational-wave observations.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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

NARCIS (Netherlands)

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

2014-01-01

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

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

Indian Academy of Sciences (India)

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

The human core of the shared socioeconomic pathways: Population scenarios by age, sex and level of education for all countries to 2100.

Science.gov (United States)

Kc, Samir; Lutz, Wolfgang

2017-01-01

This paper applies the methods of multi-dimensional mathematical demography to project national populations based on alternative assumptions on future, fertility, mortality, migration and educational transitions that correspond to the five shared socioeconomic pathways (SSP) storylines. In doing so it goes a significant step beyond past population scenarios in the IPCC context which considered only total population size. By differentiating the human population not only by age and sex-as is conventionally done in demographic projections-but also by different levels of educational attainment the most fundamental aspects of human development and social change are being explicitly addressed through modeling the changing composition of populations by these three important individual characteristics. The scenarios have been defined in a collaborative effort of the international Integrated Assessment Modeling community with the medium scenario following that of a major new effort by the Wittgenstein Centre for Demography and Global Human Capital (IIASA, OEAW, WU) involving over 550 experts from around the world. As a result, in terms of total world population size the trajectories resulting from the five SSPs stay very close to each other until around 2030 and by the middle of the century already a visible differentiation appears with the range between the highest (SSP3) and the lowest (SSP1) trajectories spanning 1.5 billion. The range opens up much more with the SSP3 reaching 12.6 billion in 2100 and SSP1 falling to 6.9 billion which is lower than today's world population.

On-line core monitoring with CORE MASTER / PRESTO

International Nuclear Information System (INIS)

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

1986-01-01

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

Defining climate change scenario characteristics with a phase space of cumulative primary energy and carbon intensity

Science.gov (United States)

Ritchie, Justin; Dowlatabadi, Hadi

2018-02-01

Climate change modeling relies on projections of future greenhouse gas emissions and other phenomena leading to changes in planetary radiative forcing. Scenarios of socio-technical development consistent with end-of-century forcing levels are commonly produced by integrated assessment models. However, outlooks for forcing from fossil energy combustion can also be presented and defined in terms of two essential components: total energy use this century and the carbon intensity of that energy. This formulation allows a phase space diagram to succinctly describe a broad range of possible outcomes for carbon emissions from the future energy system. In the following paper, we demonstrate this phase space method with the Representative Concentration Pathways (RCPs) as used in the Intergovernmental Panel on Climate Change (IPCC) Fifth Assessment Report (AR5). The resulting RCP phase space is applied to map IPCC Working Group III (WGIII) reference case ‘no policy’ scenarios. Once these scenarios are described as coordinates in the phase space, data mining techniques can readily distill their core features. Accordingly, we conduct a k-means cluster analysis to distinguish the shared outlooks of these scenarios for oil, gas and coal resource use. As a whole, the AR5 database depicts a transition toward re-carbonization, where a world without climate policy inevitably leads to an energy supply with increasing carbon intensity. This orientation runs counter to the experienced ‘dynamics as usual’ of gradual decarbonization, suggesting climate change targets outlined in the Paris Accord are more readily achievable than projected to date.

Transportation energy scenario analysis technical report No. 1: examination of four existing scenarios. [Projections for 1985, 1995, 2010, and 2025

Energy Technology Data Exchange (ETDEWEB)

Bernard, M. J.; LaBelle, S. J.; Millar, M.; Walbridge, E. W.

1978-03-01

This project aims to provide the DOE Division of Transportation Energy Conservation (TEC) with a long-range forecasting framework in which to evaluate potential changes to the U.S. Transportation system. This initial report examines four existing, but diverse, 50-year scenarios of the future. It describes the scenarios and summarizes the changes in the major transportation system variables that would occur through the year 2025 in each scenario. Projections of variables of interest to TEC are explored, including passenger or ton miles and energy consumption. Each is reported for 1985, 1995, 2010, and 2025 under four scenarios: success, moderate economic growth, energy crisis, and transformation. The philosophy of this project is that the transportation system must support future lifestyles; by examining potential future lifestyles the required transportation system changes can be deduced. The project: (a) develops a set of scenarios that span likely futures; (b) describes the lifestyles in each scenario in order; (c) determines the characteristics of the transportation system supporting those lifestyles; (d) indicates transportation technologies and policies necessary in that system; and (e) derives the energy characteristics of that system. The implications of the four existing scenarios are examined with emphasis on current TEC electric-vehicle development. This preliminary investigation will be followed by detailed-scenario building (modifying existing scenarios or developing new ones) and generation of lifestyles and transportation system demands under each of the scenarios. This work will be reported in October 1978.

Effect of RCIC Operating Conditions on the Accident Scenario in Fukushima Unit 2

International Nuclear Information System (INIS)

Kim, Sung Il; Park, Jong Hwa; Ha, Kwang Soon

2015-01-01

This study was conducted by using MELCOR 1.8.6. Fukushima unit 2 accident was analyzed using MELCOR in this study, and best estimate scenario with considering RCIC operating conditions was presented. Researches on the boiling water reactor (BWR) plant with reactor core isolation cooling (RCIC) system have been conducted. Research on the RCIC operation in Fukushima unit 2 was also conducted by Sandia National Laboratory. MELCOR analysis of the Fukushima unit 2 accident was conducted in the report and energy balance in wetwell was described by considering RCIC operation. However, the effect of RCIC operation condition on the accident scenario has not been studied. The operating conditions of RCIC system affect the pressures in wetwell and drywell, and the high pressure can make leakage path of fission product from PCV to reactor building. Thus it can be directly related with the amount of fission product which released to environment. In this study, severe accident on Fukushima unit 2 was analyzed considering the operating condition of RCIC system, and best estimated scenario was presented. In addition, the effect of RCIC turbine efficiency on the accident progression was examined. Energy balance in suppression chamber was also considered with discussion on the effect of torus room flooding level. It was found that the operating condition of RCIC turbine not only affects the variation of drywell pressure but also the amount of released fission products to environment. It was also confirmed that the RCIC turbine efficiency in the accident would be less than normal operating condition

Development of internal transport barrier scenarios at ITER-relevant high triangularity in Jet

Energy Technology Data Exchange (ETDEWEB)

Rimini, F.G.; Becoulet, M.; Ekedahl, A.; Huysmans, G.; Joffrin, E.; Litaudon, X. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Giovannozzi, E.; Tudisco, O.; Crisanti, F. [Association Euratol/ENEA/CNR sulla Fusione, Frascali, Rome (Italy); Lomas, P.J.; Alper, B.; Hawkes, N.; Parail, V.; Zastrow, K.D. [Euratom/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Baar, M. de; Vries, P. de [Association Euratom-Fom, TEC Cluster, Nieuwegein (Netherlands); La Luna, E. de [Association Euratom-Ciemat, Madrid (Spain); Saibene, G. [EFDA CSU, Garching (Germany)

2004-07-01

The development of ITB s(Internal Transport Barrier) scenarios in high triangularity discharges is of particular interest for ITER advanced tokamak operation. Previous JET experiments have shown that high triangularity favours ELM (Edge Localized Mode)-Free or type I ELMs, which inhibit long lasting ITBs. The recent experiments reported here concentrate on integrated optimisation of edge and core conditions. Edge pedestal was controlled using gas injection, Deuterium or light impurities, and plasma current ramps. Both methods yield more ITB-friendly edge pedestal conditions, varying from small type I to type III ELMs and, in extreme cases, to L-mode edge. In parallel, the conditions for triggering and sustaining a wide ITB were optimised. This plasmas have deeply reversed target current profiles with g{sub min} 3. A narrow inner ITB, located in the reversed shear region, is routinely observed. Large radius ITBs are only triggered when the input power exceeds 20-22 MW, but they do not usually survive the transition into H-mode. The best results, in terms of sustained high performance, have been obtained with Neon injection: a wide ITB is triggered during the phase with L-mode edge and survives into H-mode for about 2 s at H{sub 89}{beta}{sub N} {approx} 3.5 and {approx} 60% of the Greenwald density limit. In summary, a high triangularity scenario has been developed, which combines the desirable I characteristics of controlled edge, long lasting wide ITBs and high performance at density higher than the low triangularity JET scenarios. (authors)

Fast wave direct electron heating in advanced inductive and ITER baseline scenario discharges in DIII-D

Energy Technology Data Exchange (ETDEWEB)

Pinsker, R. I.; Jackson, G. L.; Luce, T. C.; Politzer, P. A. [General Atomics, PO Box 85608, San Diego, California 92186-5608 (United States); Austin, M. E. [University of Texas at Austin, Austin, Texas 78712 (United States); Diem, S. J.; Kaufman, M. C.; Ryan, P. M. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); Doyle, E. J.; Zeng, L. [University of California Los Angeles, Los Angeles, California 90095 (United States); Grierson, B. A.; Hosea, J. C.; Nagy, A.; Perkins, R.; Solomon, W. M.; Taylor, G. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Maggiora, R.; Milanesio, D. [Politecnico di Torino, Dipartimento di Elettronica, Torino (Italy); Porkolab, M. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Turco, F. [Columbia University, New York, New York 10027 (United States)

2014-02-12

Fast Wave (FW) heating and electron cyclotron heating (ECH) are used in the DIII-D tokamak to study plasmas with low applied torque and dominant electron heating characteristic of burning plasmas. FW heating via direct electron damping has reached the 2.5 MW level in high performance ELMy H-mode plasmas. In Advanced Inductive (AI) plasmas, core FW heating was found to be comparable to that of ECH, consistent with the excellent first-pass absorption of FWs predicted by ray-tracing models at high electron beta. FW heating at the ∼2 MW level to ELMy H-mode discharges in the ITER Baseline Scenario (IBS) showed unexpectedly strong absorption of FW power by injected neutral beam (NB) ions, indicated by significant enhancement of the D-D neutron rate, while the intended absorption on core electrons appeared rather weak. The AI and IBS discharges are compared in an effort to identify the causes of the different response to FWs.

MIV Project: Mission scenario

DEFF Research Database (Denmark)

Ravazzotti, Mariolina T.; Jørgensen, John Leif; Thuesen, Gøsta

1997-01-01

Under the ESA contract #11453/95/NL/JG(SC), aiming at assessing the feasibility of Rendez-vous and docking of unmanned spacecrafts, a msiision scenario was defined. This report describes the secquence of manouvres and task allocations for such missions.......Under the ESA contract #11453/95/NL/JG(SC), aiming at assessing the feasibility of Rendez-vous and docking of unmanned spacecrafts, a msiision scenario was defined. This report describes the secquence of manouvres and task allocations for such missions....

An Experiment on Graph Analysis Methodologies for Scenarios

Energy Technology Data Exchange (ETDEWEB)

Brothers, Alan J.; Whitney, Paul D.; Wolf, Katherine E.; Kuchar, Olga A.; Chin, George

2005-09-30

Visual graph representations are increasingly used to represent, display, and explore scenarios and the structure of organizations. The graph representations of scenarios are readily understood, and commercial software is available to create and manage these representations. The purpose of the research presented in this paper is to explore whether these graph representations support quantitative assessments of the underlying scenarios. The underlying structure of the scenarios is the information that is being targeted in the experiment and the extent to which the scenarios are similar in content. An experiment was designed that incorporated both the contents of the scenarios and analysts’ graph representations of the scenarios. The scenarios’ content was represented graphically by analysts, and both the structure and the semantics of the graph representation were attempted to be used to understand the content. The structure information was not found to be discriminating for the content of the scenarios in this experiment; but, the semantic information was discriminating.

Action research Toolkit II: The Scenario Workshop

DEFF Research Database (Denmark)

Rasmussen, Lauge Baungaard

2003-01-01

The article describes the scenario workshop within the action research paradigm. In addtion, the maina phases and the functions of the facilitator and the participants are outlined. Finally,it describes and analyses the experiences of using the scenario workshop in practice.......The article describes the scenario workshop within the action research paradigm. In addtion, the maina phases and the functions of the facilitator and the participants are outlined. Finally,it describes and analyses the experiences of using the scenario workshop in practice....

Experimental hydrate formation and gas production scenarios based on CO{sub 2} sequestration

Energy Technology Data Exchange (ETDEWEB)

Stevens, J.C.; Howard, J.J. [ConocoPhillips, Bartlesville, OK (United States). Reservoir Laboratories; Baldwin, B.A. [Green Country Petrophysics LLC, Dewey, OK (United States); Ersland, G.; Husebo, J.; Graue, A. [Bergen Univ., Bergen (Norway). Dept. of Physics and Technology

2008-07-01

Gas hydrate production strategies have focused on depressurization or thermal stimulation of the reservoir, which in turn leads to hydrate dissociation. In order to evaluate potential production scenarios, the recovery efficiency of the natural gas from hydrate must be known along with the corresponding amounts of produced water. This study focused on the exchange of carbon dioxide (CO{sub 2}) with the natural gas hydrate and the subsequent release of free methane (CH{sub 4}). Laboratory experiments that investigated the rates and mechanisms of hydrate formation in coarse-grained porous media have shown the significance of initial water saturation and salinity on forming methane hydrates. Many of the experiments were performed in a sample holder fitted with an MRI instrument for monitoring hydrate formation. Hydrate-saturated samples were subjected to different procedures to release methane. The rates and efficiency of the exchange process were reproducible over a series of initial conditions. The exchange process was rapid and efficient in that no free water was observed in the core with MRI measurements. Injection of CO{sub 2} into the whole-core hydrate-saturated pore system resulted in methane production at the outlet end. Permeability measurements on these hydrate saturated cores during hydrate formation decreased to low values, but enough for gas transport. The lower permeability values remained constant during the methane-carbon dioxide exchange process in the hydrate structure. 12 refs., 9 figs.

Investigation into fuel pin reshuffling options in PWR in-core fuel management for enhancement of efficient use of nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Daing, Aung Tharn, E-mail: atdaing@khu.ac.kr; Kim, Myung Hyun, E-mail: mhkim@khu.ac.kr

2014-07-01

Highlights: • This paper discusses an alternative option, fuel pin reshuffling for maximization of cycle energy production. • The prediction results of isotopic compositions of each burnt pin are verified. • The operating performance is analyzed at equilibrium core with fuel pin reshuffling. • The possibility of reuse of spent fuel pins for reduction of fresh fuel assemblies is investigated. - Abstract: An alternative way to enhance efficient use of nuclear fuel is investigated through fuel pin reshuffling options within PWR fuel assembly (FA). In modeling FA with reshuffled pins, as prerequisite, the single pin calculation method is proposed to estimate the isotopic compositions of each pin of burnt FA in the core-wide environment. Subsequently, such estimation has been verified by comparing with the neutronic performance of the reference design. Two scenarios are concerned, i.e., first scenario was targeted on the improvement of the uniform flux spatial distribution and on the enhancement of neutron economy by simply reshuffling the existing fuel pins in once-burnt fuel assemblies, and second one was focused on reduction of fresh fuel loading and discharged fuel assemblies with more economic incentives by reusing some available spent fuel pins still carrying enough reactivity that are mechanically sound ascertained. In scenario-1, the operating time was merely somewhat increased for few minutes when treating eight FAs by keeping enough safety margins. The scenario-2 was proved to reduce four fresh FAs loading without largely losing any targeted parameters from the safety aspect despite loss of 14 effective full power days for operation at reference plant full rated power.

Water use implications of biofuel scenarios

Science.gov (United States)

Teter, J.; Mishra, G. S.; Yeh, S.

2012-12-01

Existing studies rely upon attributional lifecycle analysis (LCA) approaches to estimate water intensity of biofuels in liters of irrigated/evapotranspiration water consumed for biofuel production. Such approaches can be misleading. From a policy perspective, a better approach is to compare differential water impacts among scenarios on a landscape scale. We address the shortcomings of existing studies by using consequential LCA, and incorporate direct and indirect land use (changes) of biofuel scenarios, marginal vs. average biofuel water use estimates, future climate, and geographic heterogeneity. We use the outputs of a partial equilibrium economic model, climate and soil data, and a process-based crop-soil-climate-water model to estimate differences in green water (GW - directly from precipitation to soil) and blue water (BW - supplied by irrigation) use among three scenarios: (1) business-as-usual (BAU), (2) Renewable Fuels Standard (RFS) mandates, and (3) a national Low Carbon Fuel Standard (LCFS) plus the RFS scenario. We use spatial statistical methods to interpolate key climatic variables using daily climate observations for the contiguous USA. Finally, we use FAO's crop model AquaCrop to estimate the domestic GW and BW impacts of biofuel policies from 2007-2035. We assess the differences among scenarios along the following metrics: (1) crop area expansion at the county level, including prime and marginal lands, (2) crop-specific and overall annual/seasonal water balances including (a) water inflows (irrigation & precipitation), (b) crop-atmosphere interactions: (evaporation & transpiration) and (d) soil-water flows (runoff & soil infiltration), in mm 3 /acre over the relevant time period. The functional unit of analysis is the BW and GW requirements of biofuels (mm3 per Btu biofuel) at the county level. Differential water use impacts among scenarios are a primarily a function of (1) land use conversion, in particular that of formerly uncropped land classes

CIEMAT analyses of transition fuel cycle scenarios

International Nuclear Information System (INIS)

Alvarez-Velarde, F.; Gonzalez-Romero, E.M.

2010-01-01

The efficient design of strategies for the long-term sustainability of nuclear energy or the phase-out of this technology is possible after the study of transition scenarios from the current fuel cycle to a future one with advanced technologies and concepts. CIEMAT has participated in numerous fuel cycle scenarios studies for more than a decade and, from some years ago, special attention has been put in the study of transition scenarios. In this paper, the main characteristics of each studied transition scenario are described. The main results and partial conclusions of each scenario are also analyzed. As general conclusions of transition studies, we highlight that the advantages of advanced technologies in transition scenarios can be obtained by countries or regions with sufficiently large nuclear parks, with a long-term implementation of the strategy. For small countries, these advantages are also accessible with an affordable cost, by means of the regional collaboration during several decades. (authors)

Scenarios of future energy intensities

International Nuclear Information System (INIS)

Anon.

1992-01-01

In this chapter, the authors present scenarios of potential change in energy intensities in the OECD countries and in the Soviet Union. These scenarios are meant to illustrate how intensities might evolve over the next 20 years given different conditions with respect to energy prices, energy-efficiency policies, and other key factors. Changes in intensity will also be affected by the rates of growth and stock turnover in each sector. They have not tried to forecast how activity levels and structure will evolve. However, the OECD scenarios assume a world in which GDP averages growth in the 2-3%/year range, with some differences among countries. For the Soviet Union, the degree and pace of intensity decline will be highly dependent on the success of the transition to a market economy; each scenario explicitly envisions a different degree of success. They have not constructed comparable scenarios for the developing countries. The scenarios presented in this chapter do not predict what will happen in the future. They believe, however, that they illustrate a plausible set of outcomes if energy prices, policies, programs, and other factors evolve as described in each case. With higher energy prices and vigorous policies and programs, intensities in the OECD countries in 2010 could be nearly 50% less on average than the level where trends seem to be point. In the former Soviet Union, a combination of rapid, successful economic reform and extra effort to improve energy efficiency might result in average intensity being nearly 40% less than in a slow reform case. And in the LDCs, a mixture of sound policies, programs, and energy pricing reform could also lead to intensities being far lower than they would be otherwise. 8 refs., 10 figs., 1 tab

Energy Efficient Routing Algorithms in Dynamic Optical Core Networks with Dual Energy Sources

DEFF Research Database (Denmark)

Wang, Jiayuan; Fagertun, Anna Manolova; Ruepp, Sarah Renée

2013-01-01

This paper proposes new energy efficient routing algorithms in optical core networks, with the application of solar energy sources and bundled links. A comprehensive solar energy model is described in the proposed network scenarios. Network performance in energy savings, connection blocking...... probability, resource utilization and bundled link usage are evaluated with dynamic network simulations. Results show that algorithms proposed aiming for reducing the dynamic part of the energy consumption of the network may raise the fixed part of the energy consumption meanwhile....

Restraint system for core elements of a reactor core

International Nuclear Information System (INIS)

Class, G.

1975-01-01

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

Energy scenarios for the 21. century

International Nuclear Information System (INIS)

Lauerman, V.

2002-01-01

The Canadian Energy Research Institute (CERI) has adopted a scenario approach in developing its energy outlook to 2025. These scenarios can be used in developing strategies to optimize opportunities and avoid the dangers that often accompany the more deterministic approach to forecasting. The scenarios are not predictions, but are instead feasible outcomes for the future. They encompass 5 major areas of uncertainty for the world energy market through 2025 that impact energy demand and fuel mix. These include the economy, technology, energy policy, resource availability, and market structure. The drivers for the energy scenarios are the environment, demography, economics, culture geopolitics and technology. The paper referred to the world energy market, the OECD energy markets, the non-OECD energy markets and the FSU energy markets under 3 energy scenarios including: (1) the material world in which the new economy drives the global economy, (2) a very dark Orwellian world in which there is a rise in terrorism and a breakdown in international cooperation, and (3) a utopia in which the major powers become increasingly serious about achieving sustainable development. 22 refs., 48 figs

Modelling of procecces in catalytic recombiners

International Nuclear Information System (INIS)

Boehm, J.

2007-01-01

In order to achieve a high degree of safety in nuclear power plants and prevent possible accident scenarios, their consequences are calculated and analysed with numeric codes. One of the most important part of nuclear safety research of hazardous incidents are development and validation of these numeric models, which are implemented into accident codes. The severe hydrogen release during a core meltdown is one of the considered scenario of performed accident analyses. One of the most important measure for the elimination of the hydrogen is catalytic recombiners. Converting the hydrogen with the atmospheric oxygen to water vapor in an exothermic reaction will prevent possible detonation of the hydrogen/air atmosphere. Within the dissertation the recombiner simulation REKO-DIREKT was developed and validated by an extensive experimental database. The performance of recombiners with regard to the conversion of the hydrogen and the temperature development is modelled. The REKO-DIREKT program is unique and has made significant revolution in research of hydrogen safety. For the first time it has been possible to show the performance of the recombiner so great in detail by using REKO-DIREKT. In the future engineers of nuclear power plants will have opportunity to have precise forecasts about the process of the possible accidents with hydrogen release. Also with presence of water vapor or with oxygen depletion which are included in the model. The major discussion of the hydrogen ignition at hot catalyst steel plates can be evaluated in the future with REKO-DIREKT more reliably than the existing used models. (orig.)

Future nuclear energy scenarios for Europe

International Nuclear Information System (INIS)

Roelofs, F.; Van Heek, A.

2010-01-01

Nuclear energy is back on the agenda worldwide. In order to prepare for the next decades and to set priorities in nuclear R and D and investment, market share scenarios are evaluated. This allows to identify the triggers which influence the market penetration of future nuclear reactor technologies. To this purpose, scenarios for a future nuclear reactor park in Europe have been analysed applying an integrated dynamic process modelling technique. Various market share scenarios for nuclear energy are derived including sub-variants with regard to the intra-nuclear options taken, e.g. introduction date of Gen-III (i.e. EPR) and Gen-IV (i.e. SCWR, HTR, FR) reactors, level of reprocessing, and so forth. The assessment was undertaken using the DANESS code which allows to provide a complete picture of mass-flow and economics of the various nuclear energy system scenarios. The analyses show that the future European nuclear park will exist of combinations of Gen-III and Gen-IV reactors. This mix will always consist of a set of reactor types each having its specific strengths. Furthermore, the analyses highlight the triggers influencing the choice between different nuclear energy deployment scenarios. In addition, a dynamic assessment is made with regard to manpower requirements for the construction of a future nuclear fleet in the different scenarios. (authors)

Implementation of an Open-Scenario, Long-Term Space Debris Simulation Approach

Science.gov (United States)

Nelson, Bron; Yang Yang, Fan; Carlino, Roberto; Dono Perez, Andres; Faber, Nicolas; Henze, Chris; Karacalioglu, Arif Goktug; O'Toole, Conor; Swenson, Jason; Stupl, Jan

2015-01-01

-track object catalog in LEO. We then use a high precision propagator to propagate all objects over the entire simulation duration. If collisions are detected, the appropriate number of debris objects are created and inserted into the simulation framework. Depending on the scenario, further objects, e.g. due to new launches, can be added. At the end of the simulation, the total number of objects above a cut-off size and the number of detected collisions provide benchmark parameters for the comparison between scenarios. The simulation approach is computationally intensive as it involves tens of thousands of objects; hence we use a highly parallel approach employing up to a thousand cores on the NASA Pleiades supercomputer for a single run. This paper describes our simulation approach, the status of its implementation, the approach to developing scenarios and examples of first test runs.

Syria and Iran: what next? Three potential scenarios; Syrie, Iran: et maintenant? Trois scenarios

Energy Technology Data Exchange (ETDEWEB)

Tertrais, Bruno [Fondation pour la recherche strategique, 4 bis rue des Patures 75016 Paris (France)

2013-10-15

Now that the United Nations resolution has been voted, the most complicated task begins: ensuring that Syria entirely and definitively gets rid of its chemical arsenal, which unfortunately has every chance of enduring. In this respect three scenarios are possible, each of which would have different repercussions for the region, notably regarding the Iranian nuclear issue, which is now being considered in a new light following Hassan Rohani's charm offensive. Yet, these consequences could also have a profound and lasting effect on the relationship between the United States and its allies. In the first scenario the agreement with Syria will be applied in a satisfactory manner, and could serve as a positive example to Iran, which might be more inclined to signing a nuclear deal. To date, Ali Khamenei, the Supreme Guide, has been wary of the ongoing negotiations, considering that it is impossible to trust the Western States whose sole objective is the downfall of the Iranian regime. Russia, buoyed by its new role in the region, would contribute in a more constructive manner to the resolution of the Iranian nuclear issue... Care must be taken, however, as Iran could just as well conclude as a result of the events of the summer of 2013, particularly from U.S. procrastination, that Western States are not willing to use force in the Middle East, and thus on the contrary draw out the negotiations without fearing the possibility of a U.S. strike. According to the second scenario, Syria will not implement the agreement, the crisis will be renewed, and will end with a Western strike, most probably without a U.N. mandate. This scenario would boost the confidence of U.S. allies in the region, who have a tendency to doubt the United States' determination to defend them. This could be seen as a warning to Iran, but the lessons that Iran would take from this scenario would undoubtedly depend on the balance of power in Tehran at the time: the moderates would push for a deal to avoid

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

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

2014-01-01

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

Composition of the low seismic velocity E' layer at the top of Earth's core

Science.gov (United States)

Badro, J.; Brodholt, J. P.

2017-12-01

Evidence for a layer (E') at the top of the outer core has been available since the '90s and while different studies suggest slightly different velocity contrasts and thicknesses, the common observation is that the layer has lower velocities than the bulk outer core (PREM). Although there are no direct measurements on the density of this layer, dynamic stability requires it to be less dense than the bulk outer core under those same pressure and temperature conditions. Using ab initio simulations on Fe-Ni-S-C-O-Si liquids we constrain the origin and composition of the low-velocity layer E' at the top of Earth's outer core. We find that increasing the concentration of any light-element always increases velocity and so a low-velocity and low-density layer (for stability) cannot be made by simply increasing light element concentration. This rules out barodiffusion or upwards sedimentation of a light phase for its origin. However, exchanging elements can—depending on the elements exchanged—produce such a layer. We evaluate three possibilities. Firstly, crystallization of a light phase from a core containing more than one light element may make such a layer, but only if the crystalizing phase is very Fe-rich, which is at odds with available phase diagrams at CMB conditions. Secondly, the E' layer may result from incomplete mixing of an early Earth core with a late impactor, depending on the light element compositions of the impactor and Earth's core, but such a primordial stratification is neither supported by dynamical models of the core nor thermodynamic models of core merger after the giant impact. The last and most plausible scenario is core-mantle chemical interaction; using thermodynamic models for metal-silicate partitioning of silicon and oxygen at CMB conditions, we show that a reaction between the core and an FeO-rich basal magma ocean can enrich the core in oxygen while depleting it in silicon, in relative amounts that produce a light and slow layer

Comparison of future energy scenarios for Denmark

DEFF Research Database (Denmark)

Kwon, Pil Seok; Østergaard, Poul Alberg

2012-01-01

Scenario-making is becoming an important tool in energy policy making and energy systems analyses. This article probes into the making of scenarios for Denmark by presenting a comparison of three future scenarios which narrate 100% renewable energy system for Denmark in 2050; IDA 2050, Climate...... Commission 2050, and CEESA (Coherent Energy and Environmental System Analysis). Generally, although with minor differences, the scenarios suggest the same technological solutions for the future such as expansion of biomass usage and wind power capacity, integration of transport sector into the other energy...

Simulationsexperimente zum Ausbreitungsverhalten von Kernschmelzen: KATS-8 bis KATS-17

International Nuclear Information System (INIS)

Eppinger, B; Fieg, G.; Massier, H.; Schuetz, W.; Stegmaier, U.; Stern, G.

2001-03-01

In future Light Water Reactors special devices (core catchers) might be required 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. Therefore, a series of experiments to investigate high temperature melt spreading on flat surfaces has been carried out 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. Spreading of oxidic and metallic melts have been performed in one- and two-dimensional geometry. Substrates were inert ceramical layer, dry concrete and concrete with a water layer of several millimeters. The influence of a shallow water layer on the surface onto the spreading behaviour has also been studied. (orig.) [de

USING UML SCENARIOS IN B2B SYSTEMS

Directory of Open Access Journals (Sweden)

A. Jakimi

2010-05-01

Full Text Available Scenarios has become a popular technique for requirements elicitation and specification building. Since scenarios capture only partial descriptions of the system behavior, an approach for scenario composition and/or integration is needed to produce more complete specifications. The Unified Modeling Language (UML, which has become a standard notation for object-oriented modeling, provides a suitable framework for scenario acquisition using Use Case diagrams and Sequence or Collaboration diagrams. In this paper, we suggest an algorithmic and tool support for composing and integrating scenarios that are represented in form of sequence diagrams. We suggest four operators (;: sequential operator, ||: concurrent operator, ?: conditional operator and  * :iteration operator to compose a set of scenarios that describe a use case of a given system. In this paper, we suggest also to apply the scenario approach to B2B systems (Business to Business. We propose to develop B2B systems as a three activities process deriving formal specifications and code skeletons from UML scenarios. Activities of this proposed process are generally automatic and are supported by a set of developed algorithms and tools.

Scenario Archetypes: Converging Rather than Diverging Themes

Directory of Open Access Journals (Sweden)

Jon P. Sadler

2012-04-01

Full Text Available Future scenarios provide challenging, plausible and relevant stories about how the future could unfold. Urban Futures (UF research has identified a substantial set (>450 of seemingly disparate scenarios published over the period 1997–2011 and within this research, a sub-set of >160 scenarios has been identified (and categorized based on their narratives according to the structure first proposed by the Global Scenario Group (GSG in 1997; three world types (Business as Usual, Barbarization, and Great Transitions and six scenarios, two for each world type (Policy Reform—PR, Market Forces—MF, Breakdown—B, Fortress World—FW, Eco-Communalism—EC and New Sustainability Paradigm—NSP. It is suggested that four of these scenario archetypes (MF, PR, NSP and FW are sufficiently distinct to facilitate active stakeholder engagement in futures thinking. Moreover they are accompanied by a well-established, internally consistent set of narratives that provide a deeper understanding of the key fundamental drivers (e.g., STEEP—Social, Technological, Economic, Environmental and Political that could bring about realistic world changes through a push or a pull effect. This is testament to the original concept of the GSG scenarios and their development and refinement over a 16 year period.

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

International Nuclear Information System (INIS)

Tobita, Yoshiharu

2013-01-01

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

Free-boundary simulations of ITER advanced scenarios

International Nuclear Information System (INIS)

Besseghir, K.

2013-06-01

The successful operation of ITER advanced scenarios is likely to be a major step forward in the development of controlled fusion as a power production source. ITER advanced scenarios raise specific challenges that are not encountered in presently-operated tokamaks. In this thesis, it is argued that ITER advanced operation may benefit from optimal control techniques. Optimal control ensures high performance operation while guaranteeing tokamak integrity. The application of optimal control techniques for ITER operation is assessed and it is concluded that robust optimisation is appropriate for ITER operation of advanced scenarios. Real-time optimisation schemes are discussed and it is concluded that the necessary conditions of optimality tracking approach may potentially be appropriate for ITER operation, thus offering a viable closed-loop optimal control approach. Simulations of ITER advanced operation are necessary in order to assess the present ITER design and uncover the main difficulties that may be encountered during advanced operation. The DINA-CH and CRONOS full tokamak simulator is used to simulate the operation of the ITER hybrid and steady-state scenarios. It is concluded that the present ITER design is appropriate for performing a hybrid scenario pulse lasting more than 1000 sec, with a flat-top plasma current of 12 MA, and a fusion gain of Q ≅ 8. Similarly, a steady-state scenario without internal transport barrier, with a flat-top plasma current of 10 MA, and with a fusion gain of Q ≅ 5 can be realised using the present ITER design. The sensitivity of the advanced scenarios with respect to transport models and physical assumption is assessed using CRONOS. It is concluded that the hybrid scenario and the steady-state scenario are highly sensitive to the L-H transition timing, to the value of the confinement enhancement factor, to the heating and current drive scenario during ramp-up, and, to a lesser extent, to the density peaking and pedestal

Free-boundary simulations of ITER advanced scenarios

Energy Technology Data Exchange (ETDEWEB)

Besseghir, K.

2013-06-15

The successful operation of ITER advanced scenarios is likely to be a major step forward in the development of controlled fusion as a power production source. ITER advanced scenarios raise specific challenges that are not encountered in presently-operated tokamaks. In this thesis, it is argued that ITER advanced operation may benefit from optimal control techniques. Optimal control ensures high performance operation while guaranteeing tokamak integrity. The application of optimal control techniques for ITER operation is assessed and it is concluded that robust optimisation is appropriate for ITER operation of advanced scenarios. Real-time optimisation schemes are discussed and it is concluded that the necessary conditions of optimality tracking approach may potentially be appropriate for ITER operation, thus offering a viable closed-loop optimal control approach. Simulations of ITER advanced operation are necessary in order to assess the present ITER design and uncover the main difficulties that may be encountered during advanced operation. The DINA-CH and CRONOS full tokamak simulator is used to simulate the operation of the ITER hybrid and steady-state scenarios. It is concluded that the present ITER design is appropriate for performing a hybrid scenario pulse lasting more than 1000 sec, with a flat-top plasma current of 12 MA, and a fusion gain of Q ≅ 8. Similarly, a steady-state scenario without internal transport barrier, with a flat-top plasma current of 10 MA, and with a fusion gain of Q ≅ 5 can be realised using the present ITER design. The sensitivity of the advanced scenarios with respect to transport models and physical assumption is assessed using CRONOS. It is concluded that the hybrid scenario and the steady-state scenario are highly sensitive to the L-H transition timing, to the value of the confinement enhancement factor, to the heating and current drive scenario during ramp-up, and, to a lesser extent, to the density peaking and pedestal

The SAFRR Tsunami Scenario

Science.gov (United States)

Porter, K.; Jones, Lucile M.; Ross, Stephanie L.; Borrero, J.; Bwarie, J.; Dykstra, D.; Geist, Eric L.; Johnson, L.; Kirby, Stephen H.; Long, K.; Lynett, P.; Miller, K.; Mortensen, Carl E.; Perry, S.; Plumlee, G.; Real, C.; Ritchie, L.; Scawthorn, C.; Thio, H.K.; Wein, Anne; Whitmore, P.; Wilson, R.; Wood, Nathan J.; Ostbo, Bruce I.; Oates, Don

2013-01-01

The U.S. Geological Survey and several partners operate a program called Science Application for Risk Reduction (SAFRR) that produces (among other things) emergency planning scenarios for natural disasters. The scenarios show how science can be used to enhance community resiliency. The SAFRR Tsunami Scenario describes potential impacts of a hypothetical, but realistic, tsunami affecting California (as well as the west coast of the United States, Alaska, and Hawaii) for the purpose of informing planning and mitigation decisions by a variety of stakeholders. The scenario begins with an Mw 9.1 earthquake off the Alaska Peninsula. With Pacific basin-wide modeling, we estimate up to 5m waves and 10 m/sec currents would strike California 5 hours later. In marinas and harbors, 13,000 small boats are damaged or sunk (1 in 3) at a cost of $350 million, causing navigation and environmental problems. Damage in the Ports of Los Angeles and Long Beach amount to $110 million, half of it water damage to vehicles and containerized cargo. Flooding of coastal communities affects 1800 city blocks, resulting in $640 million in damage. The tsunami damages 12 bridge abutments and 16 lane-miles of coastal roadway, costing $85 million to repair. Fire and business interruption losses will substantially add to direct losses. Flooding affects 170,000 residents and workers. A wide range of environmental impacts could occur. An extensive public education and outreach program is underway, as well as an evaluation of the overall effort.

Transient analyses for a molten salt fast reactor with optimized core geometry

Energy Technology Data Exchange (ETDEWEB)

Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

2015-10-15

Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

Understanding the core-halo relation of quantum wave dark matter from 3D simulations.

Science.gov (United States)

Schive, Hsi-Yu; Liao, Ming-Hsuan; Woo, Tak-Pong; Wong, Shing-Kwong; Chiueh, Tzihong; Broadhurst, Tom; Hwang, W-Y Pauchy

2014-12-31

We examine the nonlinear structure of gravitationally collapsed objects that form in our simulations of wavelike cold dark matter, described by the Schrödinger-Poisson (SP) equation with a particle mass ∼10(-22)  eV. A distinct gravitationally self-bound solitonic core is found at the center of every halo, with a profile quite different from cores modeled in the warm or self-interacting dark matter scenarios. Furthermore, we show that each solitonic core is surrounded by an extended halo composed of large fluctuating dark matter granules which modulate the halo density on a scale comparable to the diameter of the solitonic core. The scaling symmetry of the SP equation and the uncertainty principle tightly relate the core mass to the halo specific energy, which, in the context of cosmological structure formation, leads to a simple scaling between core mass (Mc) and halo mass (Mh), Mc∝a(-1/2)Mh(1/3), where a is the cosmic scale factor. We verify this scaling relation by (i) examining the internal structure of a statistical sample of virialized halos that form in our 3D cosmological simulations and by (ii) merging multiple solitons to create individual virialized objects. Sufficient simulation resolution is achieved by adaptive mesh refinement and graphic processing units acceleration. From this scaling relation, present dwarf satellite galaxies are predicted to have kiloparsec-sized cores and a minimum mass of ∼10(8)M⊙, capable of solving the small-scale controversies in the cold dark matter model. Moreover, galaxies of 2×10(12)M⊙ at z=8 should have massive solitonic cores of ∼2×10(9)M⊙ within ∼60  pc. Such cores can provide a favorable local environment for funneling the gas that leads to the prompt formation of early stellar spheroids and quasars.

Current scenario

Indian Academy of Sciences (India)

First page Back Continue Last page Overview Graphics. Current scenario. India , like other parts of the world, is also facing the problem of increase in the incidence of drug resistance in tuberculosis. Multi-drug resistance (MDR, resistance to RIF & INH) and extensively drug resistant strains (X-DR, resistance to RIF, INH, FQs ...

Maximising the Effectiveness of a Scenario Planning Process: Tips for Scenario Planners in Higher Education

Science.gov (United States)

Sayers, Nicola

2011-01-01

Scenario planning is a tool which can help organisations and people to think about, and plan for, the long-term future. In basic terms, it involves creating a number of in-depth scenarios (stories), each of which tells of a different possible future for an organisation or issue, and considering how each different future might influence…

Fiscal 2000 report of investigation. Research on effective use of technology for biomass waste and diffusion scenario of the technology; 2000 nendo biomass kei haikibutsu no yuko riyo gijutsu to sono fukyu scenario no chosa hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

Examination and proposal were made on the effective uses of technologies for biomass wastes and on the diffusion scenarios, with trial calculations done to find the maximum reducible amounts in terms of carbon dioxide. In the examination of the systems, it was focused on five technologies, namely, direct combustion of straw and rice straw/CHP/heat utilization, synthesis of gasified methanol from demolition wood, biogas/CHP from livestock feces and urine, conversion to ethanol through saccharification of used paper, and conversion to biodiesel from vegetable waste oil; the examination also covered each feature, core function, use situation inside and outside Japan, technological system for each energy conversion, etc. Further, solutions were examined and proposed concerning diffusion barriers in each scenario. Consequently, the maximum amount of CO2 reduction effect was found respectively 217 thousand t-C in the straw and rice straw, 500 thousand t-C in the demolition wood, 9.53 million t-C in the livestock feces/urine, 820 thousand t-C in the used paper, and 30 thousand t-C in the vegetable waste oil. (NEDO)

Toward Managing & Automating CyberCIEGE Scenario Definition File Creation

National Research Council Canada - National Science Library

Johns, Kenneth

2004-01-01

.... These scenarios are written CyberCIEGE Scenarios Definition Language. Unfortunately, the trade-off for flexibility, extendibility and fully customizable Scenarios is syntax complexity in the Scenarios definition language...

Heat dissipating nuclear reactor

Science.gov (United States)

Hunsbedt, A.; Lazarus, J.D.

1985-11-21

Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

International Nuclear Information System (INIS)

Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

2012-01-01

A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

What is the purpose of risk studies

International Nuclear Information System (INIS)

Birkhofer, A.

1986-01-01

Over the past few years, more or less comprehensive risk analyses have been carried out for numerous nuclear power plants, especially in the USA. The first risk studies conducted, especially the Rasmussen Study and the 'German Risk Study,' investigated sequences of accidents up to their environmental impacts. Risk studies today are focused mostly on assessments of technical safety, which is why they are extended only as far as the determination of core meltdown frequencies or radionuclide releases. Studies of this type are also called probabilistic safety analyses. (orig.) [de

Modelling of QUENCH-03 and QUENCH-06 Experiments Using RELAP/SCDAPSIM and ASTEC Codes

Directory of Open Access Journals (Sweden)

Tadas Kaliatka

2014-01-01

Full Text Available To prevent total meltdown of the uncovered and overheated core, the reflooding with water is a necessary accident management measure. Because these actions lead to the generation of hydrogen, which can cause further problems, the related phenomena are investigated performing experiments and computer simulations. In this paper, for the experiments of loss of coolant accidents, performed in Forschungszentrum Karlsruhe, QUENCH-03 and QUENCH-06 are modelled using RELAP5/SCDAPSIM and ASTEC codes. The performed benchmark allowed analysing different modelling features. The recommendations for the model development are presented.

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)

First evaluation of experimental results describing pressure buildup and hydrogen distribution, obtained by experimental series E11

International Nuclear Information System (INIS)

Holzbauer, H.; Wolf, L.; Valencia, L.

1990-01-01

The beginning of the core meltdown process in the reactor pressure vessel, with a supposed release position for hydrogen in the upper part of the containment, is simulated, and the influence of the plant geometry on the distribution process is studied. In addition, the release point is shifted to the lower containment area, additional hydrogen production owing to the interaction between concrete and melt is taken into account, and sump boiling is simulated. The experimental program covers analytical investigations in the form of blind post-calculations. (DG) [de

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).

Adaptive memory: the survival scenario enhances item-specific processing relative to a moving scenario.

Science.gov (United States)

Burns, Daniel J; Hart, Joshua; Griffith, Samantha E; Burns, Amy D

2013-01-01

Nairne, Thompson, and Pandeirada (2007) found that retention of words rated for their relevance to survival is superior to that of words encoded under numerous other deep processing conditions. They suggested that our memory systems might have evolved to confer an advantage for survival-relevant information. Burns, Burns, and Hwang (2011) suggested a two-process explanation of the proximate mechanisms responsible for the survival advantage. Whereas most control tasks encourage only one type of processing, the survival task encourages both item-specific and relational processing. They found that when control tasks encouraged both types of processing, the survival processing advantage was eliminated. However, none of their control conditions included non-survival scenarios (e.g., moving, vacation, etc.), so it is not clear how this two-process explanation would explain the survival advantage when scenarios are used as control conditions. The present experiments replicated the finding that the survival scenario improves recall relative to a moving scenario in both a between-lists and within-list design and also provided evidence that this difference was accompanied by an item-specific processing difference, not a difference in relational processing. The implications of these results for several existing accounts of the survival processing effect are discussed.

Multiple methods for multiple futures: Integrating qualitative scenario planning and quantitative simulation modeling for natural resource decision making

Science.gov (United States)

Symstad, Amy J.; Fisichelli, Nicholas A.; Miller, Brian W.; Rowland, Erika; Schuurman, Gregor W.

2017-01-01

Scenario planning helps managers incorporate climate change into their natural resource decision making through a structured “what-if” process of identifying key uncertainties and potential impacts and responses. Although qualitative scenarios, in which ecosystem responses to climate change are derived via expert opinion, often suffice for managers to begin addressing climate change in their planning, this approach may face limits in resolving the responses of complex systems to altered climate conditions. In addition, this approach may fall short of the scientific credibility managers often require to take actions that differ from current practice. Quantitative simulation modeling of ecosystem response to climate conditions and management actions can provide this credibility, but its utility is limited unless the modeling addresses the most impactful and management-relevant uncertainties and incorporates realistic management actions. We use a case study to compare and contrast management implications derived from qualitative scenario narratives and from scenarios supported by quantitative simulations. We then describe an analytical framework that refines the case study’s integrated approach in order to improve applicability of results to management decisions. The case study illustrates the value of an integrated approach for identifying counterintuitive system dynamics, refining understanding of complex relationships, clarifying the magnitude and timing of changes, identifying and checking the validity of assumptions about resource responses to climate, and refining management directions. Our proposed analytical framework retains qualitative scenario planning as a core element because its participatory approach builds understanding for both managers and scientists, lays the groundwork to focus quantitative simulations on key system dynamics, and clarifies the challenges that subsequent decision making must address.

Estimates of future climate based on SRES emission scenarios

Energy Technology Data Exchange (ETDEWEB)

Godal, Odd; Sygna, Linda; Fuglestvedt, Jan S.; Berntsen, Terje

2000-02-14

The preliminary emission scenarios in the Special Report on Emission Scenario (SRES) developed by the Intergovernmental Panel on Climate Change (IPCC), will eventually replace the old IS92 scenarios. By running these scenarios in a simple climate model (SCM) we estimate future temperature increase between 1.7 {sup o}C and 2.8 {sup o}C from 1990 to to 2100. The global sea level rise over the same period is between 0.33 m and 0.45 m. Compared to the previous IPCC scenarios (IS92) the SRES scenarios generally results in changes in both development over time and level of emissions, concentrations, radiative forcing, and finally temperature change and sea level rise. The most striking difference between the IS92 scenarios and the SRES scenarios is the lower level of SO{sub 2} emissions. The range in CO{sub 2} emissions is also expected to be narrower in the new scenarios. The SRES scenarios result in a narrower range both for temperature change and sea level rise from 1990 to 2100 compared to the range estimated for the IS92 scenarios. (author)

Long-term scenarios for global energy demand and supply. Four global greenhouse mitigation scenarios. Final report

International Nuclear Information System (INIS)

Soerensen, B.; Meibom, P.; Kuemmel, B.

1999-01-01

The scenario method is used to investigate energy demand and supply systems for the 21st century. A geographical information system (GIS) is employed to assess the spatial match between supply and demand, and the robustness of the scenario against changes in assumptions is discussed, for scenarios using fossil fuels without carbon dioxide emissions, nuclear fuels with reduced accident and proliferation risks, and renewable energy from local and from more centralised installations: The year 2050 demand scenario is based on a very high goal satisfaction in all regions of the world, for the middle UN population projection. All energy efficiency measures that are technically ready and economic today are assumed in effect by year 2050. An increased fraction of total activities are assumed to occur in non-material sectors. Technical, economic and implementation issues are discussed, including the resilience to changes in particularly demand assumptions and the type of framework that would allow energy policy to employ any of (or a mix of) the scenario options. Results are presented as average energy flows per unit of land area. This geographically based presentation method gives additional insights, particularly for the dispersed renewable energy systems, but in all cases it allows to identify the need for energy transmission and trade between regions, and to display it in a visually suggestive fashion. The scenarios are examples of greenhouse mitigation scenarios, all characterised by near-zero emissions of greenhouse gases to the atmosphere. All are more expensive than the present system, but only if the cost of the negative impacts from the current system is neglected. As options for global energy policy during the next decades, the clean fossil and the renewable energy options (possibly in combination) are the only realistic ones, because the safe nuclear option requires research and development that most likely will take longer time, if it can at all be carried

Long-term scenarios for global energy demand and supply. Four global greenhouse mitigation scenarios. Final report

Energy Technology Data Exchange (ETDEWEB)

Soerensen, B; Meibom, P [Technical Univ. of Denmark, Lyngby (Denmark); Kuemmel, B [Royal Agricultural and Veterinary Univ., Tastrup (Denmark)

1999-01-01

The scenario method is used to investigate energy demand and supply systems for the 21st century. A geographical information system (GIS) is employed to assess the spatial match between supply and demand, and the robustness of the scenario against changes in assumptions is discussed, for scenarios using fossil fuels without carbon dioxide emissions, nuclear fuels with reduced accident and proliferation risks, and renewable energy from local and from more centralised installations: The year 2050 demand scenario is based on a very high goal satisfaction in all regions of the world, for the middle UN population projection. All energy efficiency measures that are technically ready and economic today are assumed in effect by year 2050. An increased fraction of total activities are assumed to occur in non-material sectors. Technical, economic and implementation issues are discussed, including the resilience to changes in particularly demand assumptions and the type of framework that would allow energy policy to employ any of (or a mix of) the scenario options. Results are presented as average energy flows per unit of land area. This geographically based presentation method gives additional insights, particularly for the dispersed renewable energy systems, but in all cases it allows to identify the need for energy transmission and trade between regions, and to display it in a visually suggestive fashion. The scenarios are examples of greenhouse mitigation scenarios, all characterised by near-zero emissions of greenhouse gases to the atmosphere. All are more expensive than the present system, but only if the cost of the negative impacts from the current system is neglected. As options for global energy policy during the next decades, the clean fossil and the renewable energy options (possibly in combination) are the only realistic ones, because the safe nuclear option requires research and development that most likely will take longer time, if it can at all be carried

Energy scenarios for Colombia: process and content