Direct containment heating integral effects tests in geometries of European nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Meyer, Leonhard [Forschungszentrum Karlsruhe (FZK), Postfach 3640, 76021 Karlsruhe (Germany)], E-mail: meyer@iket.fzk.de; Albrecht, Giancarlo [Forschungszentrum Karlsruhe (FZK), Postfach 3640, 76021 Karlsruhe (Germany); Caroli, Cataldo [Institut de Radioprotection et de Surete Nucleaire, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Ivanov, Ivan [Technical University of Sofia, BG-1797 Sofia (Bulgaria)

2009-10-15

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P'4, the VVER-1000 and the German Konvoi plant. A high-temperature iron-alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified. The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.

Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

Energy Technology Data Exchange (ETDEWEB)

Jackson, J. D.; Li, J.; Wang, J. [The Univ., of Manchester, Manchester (United Kingdom)

2003-07-01

Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results.

Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

Science.gov (United States)

Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

2012-06-01

Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

Containment heat removal system

International Nuclear Information System (INIS)

Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

1992-01-01

This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

International Nuclear Information System (INIS)

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O.; Nichols, R.T.

1994-05-01

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report

Experiments to investigate direct containment heating phenomena with scaled models of the Zion Nuclear Power Plant in the Surtsey Test Facility

Energy Technology Data Exchange (ETDEWEB)

Allen, M.D.; Pilch, M.M.; Blanchat, T.K.; Griffith, R.O. [Sandia National Labs., Albuquerque, NM (United States); Nichols, R.T. [Ktech Corp., Albuquerque, NM (United States)

1994-05-01

The Surtsey Facility at Sandia National Laboratories (SNL) is used to perform scaled experiments that simulate hypothetical high-pressure melt ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effect of specific phenomena associated with direct containment heating (DCH) on the containment load, such as the effect of physical scale, prototypic subcompartment structures, water in the cavity, and hydrogen generation and combustion. In the Integral Effects Test (IET) series, 1:10 linear scale models of the Zion NPP structures were constructed in the Surtsey vessel. The RPV was modeled with a steel pressure vessel that had a hemispherical bottom head, which had a 4-cm hole in the bottom head that simulated the final ablated hole that would be formed by ejection of an instrument guide tube in a severe NPP accident. Iron/alumina/chromium thermite was used to simulate molten corium that would accumulate on the bottom head of an actual RPV. The chemically reactive melt simulant was ejected by high-pressure steam from the RPV model into the scaled reactor cavity. Debris was then entrained through the instrument tunnel into the subcompartment structures and the upper dome of the simulated reactor containment building. The results of the IET experiments are given in this report.

Containment for low temperature district nuclear-heating reactor

International Nuclear Information System (INIS)

He Shuyan; Dong Duo

1992-03-01

Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

Heating experiments of JT-60

International Nuclear Information System (INIS)

1987-01-01

In JT-60, after the finish of the first stage Joule experiment, the heating facilities were installed, and the heating experiment was started in August, 1986. As to neutral beam injection, the beam injection experiment at the maximum rating 20 MW carried out, and also as to RF, the injection experiment up to 1.4 MW was carried out in both ion cyclotron and low band hybrid waves. The results worthy of special mention in the heating experiment were the success in the current drive up to 1.7 MA at maximum using low band hybrid waves and the improvement of plasma confinement characteristics obtained by the compound heating of NBI and RF. In this paper, the main results of these heating experiments and their significance are explained. The JT-60 is the testing facilities for attaining the critical plasma condition by additionally heating the plasma which is generated by Joule electric discharge with NBI and RF heatings. The experimental operation cycle of the JT-60 consists of the unit cycle of two weeks, and the number of days in operation is nine days. The temperature of heated plasma rose to 70 million deg C in the 20 MW NBI heating. Hereafter, the improvement of confinement time by increasing the stored energy of plasma is attempted. (Kako, I.)

The probability of containment failure by direct containment heating in surry

International Nuclear Information System (INIS)

Pilch, M.M.; Allen, M.D.; Bergeron, K.D.; Tadios, E.L.; Stamps, D.W.; Spencer, B.W.; Quick, K.S.; Knudson, D.L.

1995-05-01

In a light-water reactor core melt accident, if the reactor pressure vessel (RPV) fails while the reactor coolant system (RCS) at high pressure, the expulsion of molten core debris may pressurize the reactor containment building (RCB) beyond its failure pressure. A failure in the bottom head of the RPV, followed by melt expulsion and blowdown of the RCS, will entrain molten core debris in the high-velocity steam blowdown gas. This chain of events is called a high-pressure melt ejection (HPME). Four mechanisms may cause a rapid increase in pressure and temperature in the reactor containment: (1) blowdown of the RCS, (2) efficient debris-to-gas heat transfer, (3) exothermic metal-steam and metal-oxygen reactions, and (4) hydrogen combustion. These processes, which lead to increased loads on the containment building, are collectively referred to as direct containment heating (DCH). It is necessary to understand factors that enhance or mitigate DCH because the pressure load imposed on the RCB may lead to early failure of the containment

CONTAIN calculations of direct containment heating in the Surry plant

International Nuclear Information System (INIS)

Williams, D.C.; Louie, D.L.Y.

1988-01-01

The draft NUREG-1150 risk analysis performed for the Surry plant identified direct containment heating (DCH) as a potentially dominant contributor to the total public risk associated with this plant. At that time, however, detailed mechanistic calculations of DCH loads were unavailable. Subsequently, a series of analyses of DCH scenarios using the CONTAIN-DCH code was performed in order to put the treatment of DCH on a firmer basis in the final draft of NUREG-1150. The present paper describes some of the results obtained for the Surry plant. A developmental model for DCH has been incorporated into CONTAIN code. This model includes mechanistic treatments of reasonably well-understood phenomena (e.g., heat and mass transfer), together with a parametric treatment of poorly understood phenomena for which mechanistic models are unavailable (e.g., debris de-entrainment from the gas stream due to debris-structure interactions). The DCH model was described in an earlier report, but the present version incorporates a number of advances, including treatment of the chemical equilibria involved in the iron-steam reaction

District heating plants in Europe: Recent experience and innovations

International Nuclear Information System (INIS)

De Comelli, G.

1992-01-01

This paper contains a critical review of recent experience and innovative features encountered in some European district heating plants. The increased application of cogeneration is pointed out, with reference to traditional, as well as, more recent technology which makes use of combined gas-steam cycles. An example of a combined gas-steam cycle is schematically described. The relevance of fluidized bed combustion and interconnection of heat distribution grids, and their consequences to the environmentally-safe and economical employment of the plants, are evidenced

Use of an integrated containment and ultimate heat sink (UHS) response approach to evaluate nuclear power plant modifications

International Nuclear Information System (INIS)

Wetzel, M.C.; Vieira, A.T.; Patton, D.C.

1994-01-01

Detailed containment and Ultimate Heat Sink (UHS) performance evaluations often are required to support major plant modifications, such as power up-rates and steam generator replacements. These UHS and containment pressure and temperature response evaluations are interrelated. Not only is the containment heat load to the UHS a factor in these evaluations, but other heat loads, such as those from the spent fuel pool, may change as a result of the plant modification and impact containment or UHS response. Our experience is that if an integrated containment/UHS response model is developed prior to the feasibility evaluations for such plant modifications, significant savings in engineering hours can be achieved. This paper presents an overview of such a front-end engineering tool that has been developed and used to support engineering evaluations. 3 refs., 2 figs

The International Heat Stress Genotype Experiment for modeling wheat response to heat: field experiments and AgMIP-Wheat multi-model simulations

DEFF Research Database (Denmark)

Martre, Pierre; Reynolds, Matthew; Asseng, Senthold

2017-01-01

The data set contains a portion of the International Heat Stress Genotype Experiment (IHSGE) data used in the AgMIP-Wheat project to analyze the uncertainty of 30 wheat crop models and quantify the impact of heat on global wheat yield productivity. It includes two spring wheat cultivars grown...... dates. Data include local daily weather data, soil characteristics and initial soil conditions, crop measurements (anthesis and maturity dates, anthesis and final total above ground biomass, final grain yields and yields components), and cultivar information. Simulations include both daily in-season...... and end-of-season results from 30 wheat models....

The International Heat Stress Genotype Experiment for Modeling Wheat Response to Heat: Field Experiments and AgMIP-Wheat Multi-Model Simulations

Science.gov (United States)

Martre, Pierre; Reynolds, Matthew P.; Asseng, Senthold; Ewert, Frank; Alderman, Phillip D.; Cammarano, Davide; Maiorano, Andrea; Ruane, Alexander C.; Aggarwal, Pramod K.; Anothai, Jakarat;

Passive heat transport in advanced CANDU containment

International Nuclear Information System (INIS)

Krause, M.; Mathew, P.M.

1994-01-01

A passive CANDU containment design has been proposed to provide the necessary heat removal following a postulated accident to maintain containment integrity. To study its feasibility and to optimize the design, multi-dimensional containment modelling may be required. This paper presents a comparison of two CFD codes, GOTHIC and PHOENICS, for multi-dimensional containment analysis and gives pressure transient predictions from a lumped-parameter and a three-dimensional GOTHIC model for a modified CANDU-3 containment. GOTHIC proved suitable for multidimensional post-accident containment analysis, as shown by the good agreement with pressure transient predictions from PHOENICS. GOTHIC is, therefore, recommended for passive CANDU containment modelling. (author)

The probability of containment failure by direct containment heating in Zion

International Nuclear Information System (INIS)

Pilch, M.M.; Yan, H.; Theofanous, T.G.

1994-12-01

This report is the first step in the resolution of the Direct Containment Heating (DCH) issue for the Zion Nuclear Power Plant using the Risk Oriented Accident Analysis Methodology (ROAAM). This report includes the definition of a probabilistic framework that decomposes the DCH problem into three probability density functions that reflect the most uncertain initial conditions (UO 2 mass, zirconium oxidation fraction, and steel mass). Uncertainties in the initial conditions are significant, but our quantification approach is based on establishing reasonable bounds that are not unnecessarily conservative. To this end, we also make use of the ROAAM ideas of enveloping scenarios and ''splintering.'' Two causal relations (CRs) are used in this framework: CR1 is a model that calculates the peak pressure in the containment as a function of the initial conditions, and CR2 is a model that returns the frequency of containment failure as a function of pressure within the containment. Uncertainty in CR1 is accounted for by the use of two independently developed phenomenological models, the Convection Limited Containment Heating (CLCH) model and the Two-Cell Equilibrium (TCE) model, and by probabilistically distributing the key parameter in both, which is the ratio of the melt entrainment time to the system blowdown time constant. The two phenomenological models have been compared with an extensive database including recent integral simulations at two different physical scales. The containment load distributions do not intersect the containment strength (fragility) curve in any significant way, resulting in containment failure probabilities less than 10 -3 for all scenarios considered. Sensitivity analyses did not show any areas of large sensitivity

The probability of containment failure by direct containment heating in zion

International Nuclear Information System (INIS)

Pilch, M.M.; Yan, H.; Theofanous, T.G.

1994-01-01

This report is the first step in the resolution of the Direct Containment Heating (DCH) issue for the Zion Nuclear Power Plant using the Risk Oriented Accident Analysis Methodology (ROAAM). This report includes the definition of a probabilistic framework that decomposes the DCH problem into three probability density functions that reflect the most uncertain initial conditions (UO 2 mass, zirconium oxidation fraction, and steel mass). Uncertainties in the initial conditions are significant, but the quantification approach is based on establishing reasonable bounds that are not unnecessarily conservative. To this end, the authors also make use of the ROAAM ideas of enveloping scenarios and open-quotes splinteringclose quotes. Two casual relations (CRs) are used in this framework: CR1 is a model that calculates the peak pressure in the containment as a function of the initial conditions, and CR2 is a model that returns the frequency of containment failure as a function of pressure within the containment. Uncertainty in CR1 is accounted for by the use of two independently developed phenomenological models, the Convection Limited Containment Heating (CLCH) model and the Two-Cell Equilibrium (TCE) model, and by probabilistically distributing the key parameter in both, which is the ratio of the melt entrainment time to the system blowdown time constant. The two phenomenological models have been compared with an extensive data base including recent integral simulations at two different physical scales (1/10th scale in the Surtsey facility at Sandia National Laboratories and 1/40th scale in the COREXIT facility at Argonne National Laboratory). The loads predicted by these models were significantly lower than those from previous parametric calculations. The containment load distributions do not intersect the containment strength curve in any significant way, resulting in containment failure probabilities less than 10 -3 for all scenarios considered

Phase Change Material Heat Sink for an ISS Flight Experiment

Science.gov (United States)

Quinn, Gregory; Stieber, Jesse; Sheth, Rubik; Ahlstrom, Thomas

2015-01-01

A flight experiment is being constructed to utilize the persistent microgravity environment of the International Space Station (ISS) to prove out operation of a microgravity compatible phase change material (PCM) heat sink. A PCM heat sink can help to reduce the overall mass and volume of future exploration spacecraft thermal control systems (TCS). The program is characterizing a new PCM heat sink that incorporates a novel phase management approach to prevent high pressures and structural deformation that often occur with PCM heat sinks undergoing cyclic operation in microgravity. The PCM unit was made using brazed aluminum construction with paraffin wax as the fusible material. It is designed to be installed into a propylene glycol and water cooling loop, with scaling consistent with the conceptual designs for the Orion Multipurpose Crew Vehicle. This paper reports on the construction of the PCM heat sink and on initial ground test results conducted at UTC Aerospace Systems prior to delivery to NASA. The prototype will be tested later on the ground and in orbit via a self-contained experiment package developed by NASA Johnson Space Center to operate in an ISS EXPRESS rack.

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

Separate effects tests on hydrogen combustion during direct containment heating events

International Nuclear Information System (INIS)

Meyer, L.; Albrecht, G.; Kirstahler, M.; Schwall, M.; Wachter, E.

2008-01-01

In the frame of severe accident research for light water reactors Forschungszentrum Karlsruhe (FZK/IKET) operates the facilities DISCO-C and DISCO-H since 1998, conceived to investigate the direct containment heating (DCH) issue. Previous DCH experiments have investigated the corium dispersion and containment pressurization during DCH in different European reactor geometries using an iron-alumina melt and steam as model fluids. The analysis of these experiments showed that the containment was pressurized by the debris-to-gas heat transfer but also to a large part by hydrogen combustion. The need was identified to better characterize the hydrogen combustion during DCH. To address this issue separate effect tests in the DISCO-H facility were conducted. These tests reproduced phenomena occurring during DCH (injection of a hot steam-hydrogen mixture jet into the containment and ignition of the air-steam-hydrogen mixture) with the exception of corium dispersion. The effect of corium particles as igniters was simulated using sparkler systems. The data will be used to validate models in combustion codes and to extrapolate to prototypic scale. Tests have been conducted in the DISCO-H facility in two steps. First a small series of six tests was done in a simplified geometry to study fundamental parameters. Then, two tests were done with a containment geometry subdivided into a subcompartment and the containment dome. The test conditions were as follows: As initial condition in the containment an atmosphere was used either with air or with a homogeneous air-steam mixture containing hydrogen concentrations between 0 and 7 mol%, temperatures around 100 C and pressure at 2 bar (representative of the containment atmosphere conditions at vessel failure). Injection of a hot steam-hydrogen jet mixture into the reactor cavity pit at 20 bar, representative of the primary circuit blow down through the vessel and hydrogen produced during this phase. The most important variables

Ion and electron heating in ICRF heating experiments on LHD

Energy Technology Data Exchange (ETDEWEB)

Saito, K. [Nagoya Univ. (Japan). Faculty of Engineering; Kumazawa, R.; Mutoh, T. [National Inst. for Fusion Science, Toki, Gifu (Japan)] [and others

2001-02-01

This paper reports on the Ion Cyclotron Range of Frequency (ICRF) heating conducted in 1999 in the 3rd experimental campaign on the Large Helical Device (LHD) with an emphasis on the optimization of the heating regime. Specifically, an exhaustive study of seven different heating regimes was carried out by changing the RF frequency relative to the magnetic field intensity, and the dependence of the heating efficiency on H-minority concentration was investigated. It was found in the experiment that both ion and electron heating are attainable with the same experimental setup by properly choosing the frequency relative to the magnetic field intensity. In the cases of both electron heating and ion heating, the power absorption efficiency depends on the minority ion concentration. An optimum minority concentration exists in the ion heating case while, in the electron heating case, the efficiency increases with concentration monotonically. A simple model calculation is introduced to provide a heuristic understanding of these experimental results. Among the heating regimes examined in this experiment, one of the ion heating regimes was finally chosen as the optimized heating regime and various high performance discharges were realized with it. (author)

Heat removal tests for pressurized water reactor containment spray by largescale facility

International Nuclear Information System (INIS)

Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

1983-01-01

Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

Heat transfer coefficient for lead matrixing in disposal containers for used reactor fuel

International Nuclear Information System (INIS)

Mathew, P.M.; Taylor, M.; Krueger, P.A.

1985-02-01

In the Canadian Nuclear Fuel Waste Management Program, metal matrices with low melting points are being evaluated for their potential to provide support for the shell of disposal containers for used fuel, and to act as an additional barrier to the release of radionuclides. The metal matrix would be incorporated into the container by casting. To study the heat transfer processes during solidification, a steady-state technique was used, involving lead as the cast metal, to determine the overall heat transfer coefficient between the lead and some of the candidate container materials. The existence of an air gap between the cast lead and the container material appeared to control the overall heat transfer coefficient. The experimental observations indicated that the surface topography of the container material influences the heat transfer and that a smoother surface results in a greater heat transfer than a rough surface. The experimental results also showed an increasing heat transfer coefficient with increasing temperature difference across the container base plates; a model developed to base-plate bending can explain the observed results

Prenatal Experiences of Containment in the Light of Bion's Model of Container/Contained

Science.gov (United States)

Maiello, Suzanne

2012-01-01

This paper explores the idea of possible proto-experiences of the prenatal child in the context of Bion's model of container/contained. The physical configuration of the embryo/foetus contained in the maternal uterus represents the starting point for an enquiry into the unborn child's possible experiences of its state of being contained in a…

Perturbative Heat Transport Experiments on TJ-II

International Nuclear Information System (INIS)

Eguilor, S.; Castejon, F.; Luna, E. de la; Cappa, A.; Likin, K.; Fernandez, A.; Tj-II, T.

2002-01-01

Heat wave experiments are performed on TJ-II stellarator plasmas to estimate both heat diffusivity and power deposition profiles. High frequency ECRH modulation experiments are used to obtain the power deposition profiles, which is observed to be wider and duller than estimated by tracing techniques. The causes of this difference are discussed in the paper. Fourier analysis techniques are used to estimate the heat diffusivity in low frequency ECRH modulation experiments. This include the power deposition profile as a new ingredient. ECHR switch on/off experiments are exploited to obtain power deposition and heat diffusivities profile. Those quantities are compared with the obtained by modulation experiments and transport analysis, showing a good agreement. (Author) 18 refs

Perturbative Heat Transport Experiments on TJ-II

Energy Technology Data Exchange (ETDEWEB)

Eguilor, S.; Castejon, F.; Luna, E. de la; Cappa, A.; Likin, K.; Fernandez, A.; Tj-II, T.

2002-07-01

Heat wave experiments are performed on TJ-II stellarator plasmas to estimate both heat diffusivity and power deposition profiles. High frequency ECRH modulation experiments are used to obtain the power deposition profiles, which is observed to be wider and duller than estimated by tracing techniques. The causes of this difference are discussed in the paper. Fourier analysis techniques are used to estimate the heat diffusivity in low frequency ECRH modulation experiments. This include the power deposition profile as a new ingredient. ECHR switch on/off experiments are exploited to obtain power deposition and heat diffusivities profile. Those quantities are compared with the obtained by modulation experiments and transport analysis, showing a good agreement. (Author) 18 refs.

High temperature experiments on a 4 tons UF6 container TENERIFE program

Energy Technology Data Exchange (ETDEWEB)

Casselman, C.; Duret, B.; Seiler, J.M.; Ringot, C.; Warniez, P.

1991-12-31

The paper presents an experimental program (called TENERIFE) whose aim is to investigate the behaviour of a cylinder containing UF{sub 6} when exposed to a high temperature fire for model validation. Taking into account the experiments performed in the past, the modelization needs further information in order to be able to predict the behaviour of a real size cylinder when engulfed in a 800{degrees}C fire, as specified in the regulation. The main unknowns are related to (1) the UF{sub 6} behaviour beyond the critical point, (2) the relationship between temperature field and internal pressure and (3) the equivalent conductivity of the solid UF{sub 6}. In order to investigate these phenomena in a representative way it is foreseen to perform experiments with a cylinder of real diameter, but reduced length, containing 4 tons of UF{sub 6}. This cylinder will be placed in an electrically heated furnace. A confinement vessel prevents any dispersion of UF{sub 6}. The heat flux delivered by the furnace will be calibrated by specific tests. The cylinder will be changed for each test.

Heat transfer in a fuel pin shipping container

International Nuclear Information System (INIS)

Ingham, J.G.

1980-01-01

Maximum cladding temperatures occur when the IDENT 1578 fuel pin shipping container is installed in the T-3 Cask. The maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 19-pin basket reaches 400 watts. Since 45% of the energy which is generated in the fuel escapes the 19-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 400/.55 = 727 watts. Similarly, the maximum allowable cladding temperature of 800 0 F is reached when the rate of energy deposited in the 40-pin basket reaches 465 watts. Since 33% of the energy which is generated in the fuel escapes the 40-pin basket without being deposited, mostly gamma energy, the maximum allowable rate of heat generation is 465/.66 = 704 watts. The IDENT 1578 fuel pin shipping container therefore meets its thermal design criteria. IDENT 1578 can handle fuel pins with a decay heat load of 600 watts while maintaining the maximum fuel pin cladding temperature below 800 0 F. The emissivities which were determined from the test results for the basket tubes and container are relatively low and correspond to new, shiny conditions. As the IDENT 1578 container is exposed to high temperatures for extended periods of time during the transportation of fuel pins, the emissivities will probably increase. This will result in reduced temperatures

Lunar heat-flow experiment

Science.gov (United States)

Langseth, M. G.

1977-01-01

The principal components of the experiment were probes, each with twelve thermometers of exceptional accuracy and stability, that recorded temperature variations at the surface and in the regolith down to 2.5 m. The Apollo 15 experiment and the Apollo 17 probes recorded lunar surface and subsurface temperatures. These data provided a unique and valuable history of the interaction of solar energy with lunar surface and the effects of heat flowing from the deep interior out through the surface of the moon. The interpretation of these data resulted in a clearer definition of the thermal and mechanical properties of the upper two meters of lunar regolith, direct measurements of the gradient in mean temperature due to heat flow from the interior and a determination of the heat flow at the Apollo 15 and Apollo 17 sites.

Heat removing device for reactor container

International Nuclear Information System (INIS)

Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

1996-01-01

A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

9 CFR 355.25 - Canning with heat processing and hermetically sealed containers; closures; code marking; heat...

Science.gov (United States)

2010-01-01

... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Canning with heat processing and hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... IDENTIFICATION AS TO CLASS, QUALITY, QUANTITY, AND CONDITION Inspection Procedure § 355.25 Canning with heat...

Heat-processing method and facility for helium-containing metal material

International Nuclear Information System (INIS)

Kato, Takahiko; Kodama, Hideyo; Matsumoto, Toshimi; Aono, Yasuhisa; Nagata, Tetsuya; Hattori, Shigeo; Kaneda, Jun-ya; Ono, Shigeki.

1996-01-01

Electric current is supplied to an objective portion of a He-containing metal material to be applied with heat processing without causing melting, to decrease the He content of the portion. Subsequently, the defect portion of the tissues of the He-containing metal is modified by heating the portion with melting. Since electric current can be supplied to the metal material in a state where the metal material is heated and the temperature thereof is elevated, an effect of further reducing the He content can be obtained. Further, if the current supply and/or the heating relative to the metal material is performed in a vacuum or inert gas atmosphere, an effect of reducing the degradation of the surface of the objective portion to be supplied with electric current can be obtained. (T.M.)

Containment loads due to direct containment heating and associated hydrogen behavior: Analysis and calculations with the CONTAIN code

International Nuclear Information System (INIS)

Williams, D.C.; Bergeron, K.D.; Carroll, D.E.; Gasser, R.D.; Tills, J.L.; Washington, K.E.

1987-05-01

One of the most important unresolved issues governing risk in many nuclear power plants involves the phenomenon called direct containment heating (DCH), in which it is postulated that molten corium ejected under high pressure from the reactor vessel is dispersed into the containment atmosphere, thereby causing sufficient heating and pressurization to threaten containment integrity. Models for the calculation of potential DCH loads have been developed and incorporated into the CONTAIN code for severe accident analysis. Using CONTAIN, DCH scenarios in PWR plants having three different representative containment types have been analyzed: Surry (subatmospheric large dry containment), Sequoyah (ice condenser containment), and Bellefonte (atmospheric large dry containment). A large number of parameter variation and phenomenological uncertainty studies were performed. Response of DCH loads to these variations was found to be quite complex; often the results differ substantially from what has been previously assumed concerning DCH. Containment compartmentalization offers the potential of greatly mitigating DCH loads relative to what might be calculated using single-cell representations of containments, but the actual degree of mitigation to be expected is sensitive to many uncertainties. Dominant uncertainties include hydrogen combustion phenomena in the extreme environments produced by DCH scenarios, and factors which affect the rate of transport of DCH energy to the upper containment. In addition, DCH loads can be aggravated by rapid blowdown of the primary system, co-dispersal of moderate quantities of water with the debris, and quenching of de-entrained debris in water; these factors act by increasing steam flows which, in turn, accelerates energy transport. It may be noted that containment-threatening loads were calculated for a substantial portion of the scenarios treated for some of the plants considered

Heat transfer from internally heated hemispherical pools

International Nuclear Information System (INIS)

Gabor, J.D.; Ellsion, P.G.; Cassulo, J.C.

1980-01-01

Experiments were conducted on heat transfer from internally heated ZnSO 4 -H 2 O pools to the walls of hemispherical containers. This experimental technique provides data for a heat transfer system that has to date been only theoretically treated. Three different sizes of copper hemispherical containers were used: 240, 280, 320 mm in diameter. The pool container served both as a heat transfer surface and as an electrode. The opposing electrode was a copper disk, 50 mm in diameter located at the top of the pool in the center. The top surface of the pool was open to the atmosphere

Design of a cavity heat pipe receiver experiment

Science.gov (United States)

Schneider, Michael G.; Brege, Mark H.; Greenlee, William J.

1992-01-01

A cavity heat pipe experiment has been designed to test the critical issues involved with incorporating thermal energy storage canisters into a heat pipe. The experiment is a replication of the operation of a heat receiver for a Brayton solar dynamic power cycle. The heat receiver is composed of a cylindrical receptor surface and an annular heat pipe with thermal energy storage canisters and gaseous working fluid heat exchanger tubes surrounding it. Hardware for the cavity heat pipe experiment will consist of a sector of the heat pipe, complete with gas tube and thermal energy storage canisters. Thermal cycling tests will be performed on the heat pipe sector to simulate the normal energy charge/discharge cycle of the receiver in a spacecraft application.

High Energy Antimatter Telescope (HEAT) Balloon Experiment

Science.gov (United States)

Beatty, J. J.

1995-01-01

This grant supported our work on the High Energy Antimatter Telescope(HEAT) balloon experiment. The HEAT payload is designed to perform a series of experiments focusing on the cosmic ray positron, electron, and antiprotons. Thus far two flights of the HEAT -e+/- configuration have taken place. During the period of this grant major accomplishments included the following: (1) Publication of the first results of the 1994 HEAT-e+/- flight in Physical Review Letters; (2) Successful reflight of the HEAT-e+/- payload from Lynn Lake in August 1995; (3) Repair and refurbishment of the elements of the HEAT payload damaged during the landing following the 1995 flight; and (4) Upgrade of the ground support equipment for future flights of the HEAT payload.

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

Comparison of CONTAIN and TCE calculations for direct containment heating of Surry

International Nuclear Information System (INIS)

Washington, K.E.; Stuart, D.S.

1996-01-01

This paper presents the results of several CONTAIN code calculations used to model direct containment heating (DCH) loads for the Surry plant. The results of these calculations are compared with the results obtained using the two-cell equilibrium (TCE) model for the same set of initial and boundary conditions. This comparison is important because both models have been favorably validated against the available DCH database, yet there are potentially important modeling differences. The comparisons are to quantitatively assess the impact of these differences. A major conclusion of this study is that, for the accident conditions studied and for a broad range of sensitivity cases, the peak pressures predicted by both TCE and CONTAIN are well below the failure pressure for the Surry containment. (orig.)

Reflooding experiments on a 49-rod cluster containing a long 90% blockage

International Nuclear Information System (INIS)

Pearson, K.G.; Cooper, C.A.; Jowitt, D.; Kinneir, J.H.

1983-01-01

A series of reflooding experiments was performed on a model fuel assembly, containing a very severe partial blockage, in the THETIS rig. The assembly comprised 49 full length, electrically heated fuel rod simulators and the blockage was created by attaching thin-walled, preformed swellings to a group of 16 rods. Results are presented for single phase and forced reflooding experiments. The most important findings relate to the improvements in heat transfer created by spacer grids and the nature of the heat transfer processes within the blockage. Spacer grids are shown to improve heat transfer by increasing turbulence and also, when wet, by cooling the steam flowing through them. Liquid penetration evidently deteriorates as the rewetting front approaches the blockage, allowing the steam through the blockage to superheat strongly and giving rise to a late peak in cladding temperature. At low reflooding rates there is a temperature penalty associated with the blockage which becomes increasingly larger as the reflooding rate is reduced. The adequacy of cooling in this very severe blockage becomes questionable when the reflooding rate falls to about 2cm/s. (U.K.)

Study on solar chemical heat pump system. Basic experiment on falling film reaction for dehydrogenation of 2-propanol; Solar chemical heat pump no kenkyu. 2-propanol bunkai hanno ni okeru ryuka ekimakushiki hanno jikken

Energy Technology Data Exchange (ETDEWEB)

Doi, T; Ando, Y; Tanaka, T; Takashima, T [Electrochemical Laboratory, Tsukuba (Japan); Nomura, T; Kamoshida, J [Shibaura Institute of Technology, Tokyo (Japan)

1996-10-27

An experiment and the examination were carried out in order to elucidate the optimum conditions in the falling liquid film reaction method, in the conversion of solar energy using the decomposition reaction of 2-propanol. The device for the experiment was constituted of a reaction container, tubular pump, cooling pipe, sampling container for effluent from the upper and lower part of the reaction container, and gas burette. Examined in the experiment were various factors such as a fibrous activated carbon (catalyst support), ratio for carrying catalyst, catalytic composition and heating temperature. In the experiment, with the temperature inside the reaction container fully stabilized under prescribed conditions, measurement was done for the hydrogen generation by the gas burette for 10 minutes as well as for the sampling of effluent. The experiment revealed that the heat utilization ratio reached the maximum of about 27% when the heating temperature was 90{degree}C using a catalyst with the ratio of RU and Pt 1 to 1 and the ratio for carrying catalyst 10wt%, so that a great improvement was obtained in the heat utilization ratio at a low temperature. Also obtained was a large inversion ratio of about 15%. 4 refs., 6 figs., 5 tabs.

Assessment of GOTHIC and TRACE codes against selected PANDA experiments on a Passive Containment Condenser

Energy Technology Data Exchange (ETDEWEB)

Papini, Davide, E-mail: davide.papini@psi.ch; Adamsson, Carl; Andreani, Michele; Prasser, Horst-Michael

2014-10-15

Highlights: • Code comparison on the performance of a Passive Containment Condenser. • Simulation of separate effect tests with pure steam and non-condensable gases. • Role of the secondary side and accuracy of pool boiling models are discussed. • GOTHIC and TRACE predict the experimental performance with slight underestimation. • Recirculatory flow pattern with injection of light non-condensable gas is inferred. - Abstract: Typical passive safety systems for ALWRs (Advanced Light Water Reactors) rely on the condensation of steam to remove the decay heat from the core or the containment. In the present paper the three-dimensional containment code GOTHIC and the one-dimensional system code TRACE are compared on the calculation of a variety of phenomena characterizing the response of a passive condenser submerged in a boiling pool. The investigation addresses the conditions of interest for the Passive Containment Cooling System (PCCS) proposed for the ESBWR (Economic Simplified Boiling Water Reactor). The analysis of selected separate effect tests carried out on a PCC (Passive Containment Condenser) unit in the PANDA large-scale thermal-hydraulic facility is presented to assess the code predictions. Both pure steam conditions (operating pressure of 3 bar, 6 bar and 9 bar) and the effect on the condensation heat transfer of non-condensable gases heavier than steam (air) and lighter than steam (helium) are considered. The role of the secondary side (pool side) heat transfer on the condenser performance is examined too. In general, this study shows that both the GOTHIC and TRACE codes are able to reasonably predict the heat transfer capability of the PCC as well as the influence of non-condensable gas on the system. A slight underestimation of the condenser performance is obtained with both codes. For those tests where the experimental and simulated efficiencies agree better the possibility of compensating errors among different parts of the heat transfer

Heat removing device for nuclear reactor container facility

Energy Technology Data Exchange (ETDEWEB)

Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

1994-09-30

A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

Experience and Prospects of Nuclear Heat Application

International Nuclear Information System (INIS)

Woite, G.; Konishi, T.; Kupitz, J.

1998-01-01

Relevant technical characteristics of nuclear reactors and heat application facilities for district heating, process heat and seawater desalination are presented and discussed. The necessity of matching the characteristics of reactors and heat applications has consequences for their technical and economic viability. The world-wide operating experience with nuclear district heating, process heating, process heat and seawater desalination is summarised and the prospects for these nuclear heat applications are discussed. (author)

Heat transfer simulation and retort program adjustment for thermal processing of wheat based Haleem in semi-rigid aluminum containers.

Science.gov (United States)

Vatankhah, Hamed; Zamindar, Nafiseh; Shahedi Baghekhandan, Mohammad

2015-10-01

A mixed computational strategy was used to simulate and optimize the thermal processing of Haleem, an ancient eastern food, in semi-rigid aluminum containers. Average temperature values of the experiments showed no significant difference (α = 0.05) in contrast to the predicted temperatures at the same positions. According to the model, the slowest heating zone was located in geometrical center of the container. The container geometrical center F0 was estimated to be 23.8 min. A 19 min processing time interval decrease in holding time of the treatment was estimated to optimize the heating operation since the preferred F0 of some starch or meat based fluid foods is about 4.8-7.5 min.

FFTF primary heat transport system heating, ventilating and air conditioning system experience

International Nuclear Information System (INIS)

Umek, A.M.; Hicks, D.F.; Schweiger, D.L.

1981-01-01

FFTF cools its primary/in-containment sodium equipment cells by means of a forced nitrogen cooling system which exchanges heat with a water-glycol system. The nitrogen cooling system is also used to maintain an inert gas atmosphere in the cells containing sodium equipment. Sodium Piping and Components have installed electrical resistance heaters to maintain a minimum sodium temperature and stainless steel jacketed mineral insulation to reduce heat loss. Design features and test results of a comprehensive redesign of the HVAC and insulation system required to support long-term nuclear operations are discussed

MELCOR 1.8.2 Assessment: IET direct containment heating tests

Energy Technology Data Exchange (ETDEWEB)

Kmetyk, L.N.

1993-10-01

MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part of an ongoing assessment program, the MELCOR computer code has been used to analyze several of the IET direct containment heating experiments done at 1:10 linear scale in the Surtsey test facility at Sandia and at 1:40 linear scale in the corium-water thermal interactions (CWTI) COREXIT test facility at Argonne National Laboratory. These MELCOR calculations were done as an open post-test study, with both the experimental data and CONTAIN results available to guide the selection of code input. Basecase MELCOR results are compared to test data in order to evaluate the new HPME DCH model recently added in MELCOR version 1.8.2. The effect of various user-input parameters in the HPME model, which define both the initial debris source and the subsequent debris interaction, were investigated in sensitivity studies. In addition, several other non-default input modelling changes involving other MELCOR code packages were required in our IET assessment analyses in order to reproduce the observed experiment behavior. Several calculations were done to identify whether any numeric effects exist in our DCH IET assessment analyses.

Heat transfer properties of organic coolants containing high boiling residues

International Nuclear Information System (INIS)

Debbage, A.G.; Driver, M.; Waller, P.R.

1964-01-01

Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

Heating and cooling device for use in the vacuum container of a thermonuclear device

International Nuclear Information System (INIS)

Morita, Hiroaki; Onozuka, Masanori; Fukui, Hiroshi.

1986-01-01

Purpose: To prevent the generation of great temperature difference within a hollow doughnuts-shaped space of the torus vacuum container of a tokamak type thermonuclear reactor, as well as effectively eliminate the local injection of heat to the vacuum container. Constitution: A hollow doughnuts-like space is formed between the inner wall and the double outer wall of a vacuum container main body, which is divided into a plurality of regions by partition plates extended in the toroidal direction. An input/output header is disposed in adjacent with each of the partition plates for inputting/outputting heat medium. Further, heat medium inlet/outlets are disposed to define two flow channels on every one-half circumference. This enables to reduce the temperature difference of the heat medium between the inlet and the outlet by the shortening of the flow channel length and heating or cooling can be performed without causing unevenness in the temperature distribution of the vacuum container. (Horiuchi, T.)

Additional heating experiments of FRC plasmas

International Nuclear Information System (INIS)

Okada, S.; Asai, T.; Kodera, F.; Kitano, K.; Suzuki, T.; Yamanaka, K.; Kanki, T.; Inomoto, M.; Yoshimura, S.; Okubo, M.; Sugimoto, S.; Ohi, S.; Goto, S.

2001-01-01

Additional heating experiments of neutral beam (NB) injection and application of low frequency wave on a plasma with extremely high averaged beta value of about 90% - a field reversed configuration (FRC) plasma - are carried out on the FRC Injection experiment (FIX) apparatus. These experiments are made possible by translating the FRC plasma produced in a formation region of a theta pinch to a confinement region in order to secure better accessibility to heating facilities and to control plasma density. By appropriate choice of injection geometry and the mirror ratio of the confinement region, the NB with the energy of 14keV and the current of 23A is enabled to be injected into the FRC in the solenoidal confining field of only 0.04-0.05T. Confinement is improved by this experiment. Ion heating is observed by the application of low frequency (80kHz ; about 1/4 of the ion gyro frequency) compressional wave. A shear wave, probably mode converted from the compressional wave, is detected to propagate axially. (author)

An Experiment on Heat Recovery Performance Improvements in Well-Water Heat-Pump Systems for a Traditional Japanese House

Directory of Open Access Journals (Sweden)

Chiemi Iba

2018-04-01

Full Text Available Concerns about resource depletion have prompted several countries to promote the usage of renewable energy, such as underground heat. In Japan, underground heat-pump technology has begun to be utilized in large-scale office buildings; however, several economic problems are observed to still exist, such as high initial costs that include drilling requirements. Further, most of the traditional dwellings “Kyo-machiya” in Kyoto, Japan have a shallow well. This study intends to propose an effective ground-source heat-pump system using the well water from a “Kyo-machiya” home that does not contain any drilling works. In previous research, it was depicted that the well-water temperature decreases as the heat pump (HP is operated and that the heat extraction efficiency steadily becomes lower. In this study, an experiment is conducted to improve efficiency using a drainage pump. Based on the experimental results, the effect of efficiency improvement and the increase in the electric power consumption of the drainage pump are examined. It is indicated that short-time drainage could help to improve efficiency without consuming excessive energy. Thus, continuous use of the heat pump becomes possible.

CEC thermal-hydraulic benchmark exercise on Fiploc verification experiment F2 in Battelle model containment long-term heat-up phase. Results for phase I

International Nuclear Information System (INIS)

Fischer, K.; Schall, M.; Wolf, L.

1991-01-01

The major objective of the F2 experiment was to investigate the thermal-hydraulic long-term phenomena with special emphasis on natural convection phenomena in a loop-type geometry affected by variations of steam and air injections at different locations as well as dry energy supply into various compartments. The open post-test exercise is being performed in two consecutive phases, with Phase I covering the initial long-term heat-up phase. The exercise received widespread international attention with nine organizations from six European countries participating with seven different computer codes (FUMO, Jericho2, Fiploc, Wavco, Contain, Melcor, Cobra/Fathoms). These codes cover a broad spectrum of presently known European computational tools in severe accident containment analyses. The participants used either the specified mass flow or pressure control boundary conditions. Some exercised their codes for both. In total, 14 different computations were officially provided by the participants indicating strong interests and cooperative efforts by various institutions

Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)

2015-05-15

Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by

Transition to chaos in a square enclosure containing internal heat sources

Energy Technology Data Exchange (ETDEWEB)

Baytas, A.C. [Institute For Nuclear Energy, Istanbul (Turkey)

1995-09-01

A numerical investigation is performed to study the transition from steady to chaotic flow of a fluid confined in a two-dimensional square cavity. The cavity has rigid walls of constant temperature containing uniformly distributed internal heat source. Effects of the Rayleigh number of flow and heat transfer rates are studied. In addition to, same problem is solved for sinusoidally changing internal heat source to show its effect on the flow model and heat transfer of the enclosures. Details of oscillatory solutions and flow bifurcations are presented.

Heat transfer in underground heating experiments in granite, Stipa, Sweden

International Nuclear Information System (INIS)

Chan, T.; Javandel, I.; Witherspoon, P.A.

1980-04-01

Electrical heater experiments have been conducted underground in granite at Stripa, Sweden, to investigate the effects of heating associated with nuclear waste storage. Temperature data from these experiments are compared with closed-form and finite-element solutions. Good agreement is found between measured temperatures and both types of models, but especially for a nonlinear finite-element heat conduction model incorporating convective boundary conditions, measured nonuniform initial rock temperature distribution, and temperature-dependent thermal conductivity. In situ thermal properties, determined by least-squares regression, are very close to laboratory values. A limited amount of sensitivity analysis is undertaken

Effect of heat treatment duration on phase separation of sodium borosilicate glass, containing copper

International Nuclear Information System (INIS)

Shejnina, T.G.; Gutner, S.Kh.; Anan'in, N.I.

1989-01-01

The effect of heat treatment duration on phase separation of sodium borosilicate (SBS) glass, containing copper is studied. It is stated that phase separation close to equilibrium one is attained under 12 hours of heat treatment of SBS glass containing copper

Condensation heat transfer with noncondensable gas for passive containment cooling of nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Leonardi, Tauna [Schlumberger, 14910 Airline Rd., Rosharon, TX 77583 (United States)]. E-mail: Tleonardi@slb.com; Ishii, Mamoru [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)]. E-mail: Ishii@ecn.purdue.edu

2006-09-15

Noncondensable gases that come from the containment and the interaction of cladding and steam during a severe accident deteriorate a passive containment cooling system's performance by degrading the heat transfer capabilities of the condensers in passive containment cooling systems. This work contributes to the area of modeling condensation heat transfer with noncondensable gases in integral facilities. Previously existing correlations and models are for the through-flow of the mixture of steam and the noncondensable gases and this may not be applicable to passive containment cooling systems where there is no clear passage for the steam to escape. This work presents a condensation heat transfer model for the downward cocurrent flow of a steam/air mixture through a condenser tube, taking into account the atypical characteristics of the passive containment cooling system. An empirical model is developed that depends on the inlet conditions, including the mixture Reynolds number and noncondensable gas concentration.

Research on heat and mass transfer model for passive containment cooling system

International Nuclear Information System (INIS)

Jiang Xiaowei; Yu Hongxing; Sun Yufa; Huang Daishun

2013-01-01

Different with the traditional dry style containment design without external cooling, the PCCS design increased the temperature difference between the wall and the containment atmosphere significantly, and also the absolute temperature of the containment surfaces will be lower, affecting properties relevant in the condensation process. A research on the heat and mass transfer model has been done in this paper, especially the improvement on the condensation and evaporation model in the presence of noncondensable gases. Firstly, the Peterson's diffusion layer model was proved to equivalent to the stagnant film model adopted by CONTAIN code using the Clausius-Clapeyron equation, then a factor which can be used to stagnant film model was derived from the comparison between the Y.Liao's generalized diffusion layer model and the Peterson's diffusion layer model. Finally, the model in CONTAIN code used to compute the condensation and evaporation mass flux was modified using the factor, and the Wisconsin condensation tests and Westinghouse film evaporation on heated plate tests were simulated which had proved the improved model can predict more closer value of the heat and mass transfer coefficient to experimental value than original model. (authors)

Dynamic response of the target container under pulsed heating

Energy Technology Data Exchange (ETDEWEB)

Liping Ni [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1997-09-01

The structural mechanics of a liquid target container for pulsed spallation sources have been simulated using both a commercial code and a PSI-developed program. Results from the transient thermal-structural analysis showed that, due to inertia effects, the dynamic stress in the target container is contributed mainly from direct heating in the initial time stage, and later from the pressure wave in the target liquid once it reaches the wall. (author) figs., tab., refs.

The inaccuracy of heat transfer characteristics for non-insulated and insulated spherical containers neglecting the influence of heat radiation

International Nuclear Information System (INIS)

Wong, King-Leung; Salazar, Jose Luis Leon; Prasad, Leo; Chen, Wen-Lih

2011-01-01

In this investigation, the differences of heat transfer characteristics for insulated and non-insulated spherical containers between considering and neglecting the influence of heat radiation are studied by the simulations in some practical situations. It is found that the heat radiation effect cannot be ignored in conditions of low ambient convection heat coefficients (such ambient air) and high surface emissivities, especially for the non-insulated and thin insulated cases. In most practical situations when ambient temperature is different from surroundings temperature and the emissivity of insulation surface is different from that of metal wall surface, neglecting heat radiation will result in inaccurate insulation effect and heat transfer errors even with very thick insulation. However, the insulation effect considering heat radiation will only increase a very small amount after some dimensionless insulated thickness (such insulation thickness/radius ≥0.2 in this study), thus such dimensionless insulated thickness can be used as the optimum thickness in practical applications. Meanwhile, wrapping a material with low surface emissivity (such as aluminum foil) around the oxidized metal wall or insulation layer (always with high surface emissivity) can achieve very good insulated effect for the non-insulated or thin insulated containers.

Cornish heat transfer experiment - final report

International Nuclear Information System (INIS)

Bourke, P.J.; Hodgkinson, D.P.

1985-01-01

The transfer of heat released in an in-site heating experiment simulating high level radioactive waste packages in granite in Cornwall has been found to be mainly by conduction but some appreciable convection does occur. Interim analysis of the data suggests that the latter may account for about 20% of the total. (author)

Mobile heat storage containers and their transport by rail or road

Energy Technology Data Exchange (ETDEWEB)

Goldenberg, Philipp

2013-10-15

Mobile heat storage containers are capable of making a contribution to the meaningful use of energy which is needed for use at a location other than where it originates. The study presented in this report outlines the technology of mobile heat storage and analyses an example of its transport by rail or road. (orig.)

The prediction of direct containment heating

International Nuclear Information System (INIS)

Yan, H.; California Univ., Santa Barbara, CA; Theofanous, T.G.; California Univ., Santa Barbara, CA

1996-01-01

A simple analytical model is proposed and shown to capture the essence of the direct containment heating phenomenon. The model is based on assuming thermal/chemical equilibrium in the melt dispersal (flow) process, and separation of the melt out of this 'equilibrium steam' in the intermediate compartment. The model reveals a natural scale (hence named the 'DCH scale') for the DCH phenomenon, and the results are in very good agreement with the integral effects tests series. On this basis, reactor predictions can be made quite simply, provided that the DCH scale for the particular condition of interest is known. This prediction of DCH scale is also addressed by a scaling approach that is shown to be consistent with the experimental data. Finally, reactor predictions (of DCH loads) are also included in generalized terms convenient for use under a wide variety of conditions. In general, the results appear to be well within the structural capability of large dry containments. (orig.)

TENERIFE program: high temperature experiments on A 4 tons UF6 container

International Nuclear Information System (INIS)

Casselman, C.; Duret, B.; Seiler, J.M.; Ringot, C.; Warniez, P.; Wataru, M.; Shiomi, S.; Ozaki, S.; Yamakawa, H.

1993-01-01

To know the input of the future thermo-mechanical code, we have to get a better understanding of the thermo-physical evolution of the UF 6 which pressurizes the container. This evolution is function of: a) the heat transfer rate from the fire to the container b) the UF 6 behaviour in the container. These tests are essentially analytical at simulated fire temperatures of between 800 and 1000degC. They use a representative mass of UF 6 (around 4 tons). The tests will not seek to rupture the test container which has a diameter equal to the 48Y container, but shorter length. These tests carried out in realistic conditions (typical thermal gradient at the wall, characteristic period for UF 6 internal mass transfer) should make possible to improve knowledge of two fundamental phenomena: 1) vaporization of UF 6 on contact with the heated wall (around 400degC), a phenomenon which controls the container internal pressurization kinetic, 2) the equivalent conductivity of solid UF 6 , a phenomenon which is linked to the heat transfer by UF 6 vaporization-condensation through the solid's porosities and which depends on the diameter of the container. In addition, they will allow the influence of other parameters to be studied, such as UF 6 container filling mode or the mechanical characteristics of the container material. A UF 6 container fitted with instruments (wall temperature, UF 6 temperature, pressure) is heated by a rapid heat transient in a radiating furnace where the temperature and thermal power supplied can be measured. The test continues until pre-established thresholds have been reached: 1) strain threshold measured on the container surface (strain gauges positioned on the outside), 2) maximum temperature threshold of UF 6 , 3) container internal pressure threshold. (J.P.N.)

On the possible role of thermal radiation in containment thermal–hydraulics experiments by the example of CFD analysis of TOSQAN T114 air–He test

Energy Technology Data Exchange (ETDEWEB)

Filippov, A.S.; Grigoryev, S.Yu. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Moscow Institute of Physics and Technology (Russian Federation); Tarasov, O.V. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation)

2016-12-15

Highlights: • Neglecting by heat radiation in simulation of containment tests may cause discrepancies. • To show that, heat exchange in T114 air-helium test was analyzed in different ways. • Effect of thermal radiation on local temperature was numerically obtained in air with ∼1% steam content. • Model of gas-structure heat exchange in containment should include heat radiation. - Abstract: One of the experiments of ERCOSAM–SAMARA (E–S) projects (TOSQAN T114) is examined from the viewpoint of the radiative heat transfer (RHT) contribution to the overall heat exchange. E–S projects and T114 test were focused on investigation of light gas stratification in severe accident containment atmosphere and stratification break-up after the activation of mitigation systems. The first from two phases of T114 test is considered during which helium is quasistatically injected into the upper part of the TOSQAN vessel having isothermal walls and initially filled by air. The developing free convection removes most of the heat acquired, but not all. Thus stable local deviations in calculated temperatures were obtained in simulations that were interpreted as the deficiencies of the physical heat-transfer model. The modeling of RHT was included in full CFD simulation that resulted in a better agreement in local temperatures. The results of comparative calculations performed without/with RHT modeling are described in the paper. The RHT model implemented in the used CFD code (ANSYS FLUENT) was tested on known analytical solutions. The RHT contribution in T114 test was also estimated analytically to demonstrate independently that it may be noticeable in this experiment. The same estimations may be valid for stagnant zones of severe accident containment. All that shows the need in further detailing of the role of RHT in gas-structure heat exchange: as for interpretation of some containment tests performed in pressure vessel as for containment modeling.

Theoretical Support of Heat Exchanger Experiments of the EU-CONGA Project

International Nuclear Information System (INIS)

Herranz, L. E.; Lopez Jimenez, J.; Munoz-Cobo, J. L.; Palomo, M. J.

1999-01-01

In this report the work carried out within the Work Package 5 of the CONGA project under the auspices of the European Union has been presented. Primarily focused on studying from a theoretical perspective the degradation of heat exchangers to be used in next generation of European reactor containments under accident conditions, and particularly the effect of aerosols, the objective has been met quite satisfactorily and the results can be summed up in three specific items: - A mathematical model of a mechanistic nature that has been encapsulated into a FORTRAN code (HTCFOUL) capable of simulating condensation heat transfer to a horizontal finned tube internally cooled. - A theoretical correlation depending upon non-dimensional variables and numbers which embodies most of the HTCFOUL physics and gives results not beyond 20% of actual HTCFOUL estimates. - A reasonable interpretation of the major measurements and observations obtained in the heat exchanger experiments performed within the Work Package 2 of the CONGA project. (Author) 55 refs

ICRF heating experiments on JIPP T-II

International Nuclear Information System (INIS)

Ichimura, M.; Fujita, J.; Hirokura, S.

1983-10-01

Data of JIPP T-II ICRF heating experiments are presented. The experiment covers three typical cases: the low concentration hydrogen minority case, the high concentration hydrogen minority case, and the 3 He minority case. The best heating efficiency is obtained for the 3 Heminority case. It is shown through power balance analysis that the two H-minority cases are different in the wave energy deposition profile. The difference is explained by the presence of local cavity mode for the high concentration minority case. The ion temperature stops rising at the power density level of 0.65 W/cm 3 . An analytic solution of the Fokker-Planck equation is derived to interpret the deterioration of heating efficiency. (author)

Combined Steady-State and Dynamic Heat Exchanger Experiment

Science.gov (United States)

Luyben, William L.; Tuzla, Kemal; Bader, Paul N.

2009-01-01

This paper describes a heat-transfer experiment that combines steady-state analysis and dynamic control. A process-water stream is circulated through two tube-in-shell heat exchangers in series. In the first, the process water is heated by steam. In the second, it is cooled by cooling water. The equipment is pilot-plant size: heat-transfer areas…

Lower hybrid heating experiment in JFT-2 tokamak

International Nuclear Information System (INIS)

Uchara, K.; Nagashima, T.

1982-01-01

Lower hybrid heating experiments in JFT-2 are reviewed. Good maintenance and controlling of the coupling structure are very important in the injection of RF power before heating experiments. Accessibility of waves and the existence of the mode conversion region are necessary for ion heating in the main plasma. Parametric instabilities which may bring undesirable power deposition are suppressed by enough electron heating in the boundary region. Optimizing the Nsub(z) spectrum and the improvement of the plasma confinement may lead the electron heating in the high density region. Current generation by use of quasi-linear Landau damping is confirmed and is suggested to bring the improvement of plasma confinement. High power and long pulse klystrons may be expected to open a frontier toward a stational reactor plasma in tokamaks. (author)

Fundamental experiment of potassium heat exchanger using principle of heat pipe

International Nuclear Information System (INIS)

Sumida, Isao; Kotani, Koichi

1976-01-01

In order to provide compact and reliable sodium equipments including a steam generator, performance tests are conducted with a potassium heat exchanger, which is featured by the separate construction of primary and secondary coolant systems. A small amount of potassium plays a role as an intermediate media of heat transportation between these two coolant systems. Heat is transferred by evaporation and condensation of potassium on the surface of the primary and the secondary coolant pipings, respectively. The tests are performed in the temperature range of 200 -- 300 0 C and the maximum heat transfer reaches 1.3kW (heat transfer rate at the primary heating source: 8.6W/cm 2 at 300 0 C). The experimental results are analyzed by using Langmuir's and Schrage's equation and close agreement between experiment and theory is obtained. (auth.)

Heat transfer fluids containing nanoparticles

Science.gov (United States)

Singh, Dileep; Routbort, Jules; Routbort, A.J.; Yu, Wenhua; Timofeeva, Elena; Smith, David S.; France, David M.

2016-05-17

A nanofluid of a base heat transfer fluid and a plurality of ceramic nanoparticles suspended throughout the base heat transfer fluid applicable to commercial and industrial heat transfer applications. The nanofluid is stable, non-reactive and exhibits enhanced heat transfer properties relative to the base heat transfer fluid, with only minimal increases in pumping power required relative to the base heat transfer fluid. In a particular embodiment, the plurality of ceramic nanoparticles comprise silicon carbide and the base heat transfer fluid comprises water and water and ethylene glycol mixtures.

Using containment analysis to improve component cooling water heat exchanger limits

International Nuclear Information System (INIS)

Da Silva, H.C.; Tajbakhsh, A.

1995-01-01

The Comanche Peak Steam Electric Station design requires that exit temperatures from the Component Cooling Water Heat Exchanger remain below 330.37 K during the Emergency Core Cooling System recirculation stage, following a hypothetical Loss of Coolant Accident (LOCA). Due to measurements indicating a higher than expected combination of: (a) high fouling factor in the Component Cooling Water Heat Exchanger with (b) high ultimate heat sink temperatures, that might lead to temperatures in excess of the 330.37 K limit, if a LOCA were to occur, TUElectric adjusted key flow rates in the Component Cooling Water network. This solution could only be implemented with improvements to the containment analysis methodology of record. The new method builds upon the CONTEMPT-LT/028 code by: (a) coupling the long term post-LOCA thermohydraulics with a more detailed analytical model for the complex Component Cooling Water Heat Exchanger network and (b) changing the way mass and energy releases are calculated after core reflood and steam generator energy is dumped to the containment. In addition, a simple code to calculate normal cooldowns was developed to confirm RHR design bases were met with the improved limits

Lower hybrid heating experiments in tokamaks: an overview

International Nuclear Information System (INIS)

Porkolab, M.

1985-10-01

Lower hybrid wave propagation theory relevant to heating fusion grade plasmas (tokamaks) is reviewed. A brief discussion of accessibility, absorption, and toroidal ray propagation is given. The main part of the paper reviews recent results in heating experiments on tokamaks. Both electron and ion heating regimes will be discussed. The prospects of heating to high temperatures in reactor grade plasmas will be evaluated

Fabrication experiments for large helix heat exchangers

International Nuclear Information System (INIS)

Burgsmueller, P.

1978-01-01

The helical tube has gained increasing attention as a heat transfer element for various kinds of heat exchangers over the last decade. Regardless of reactor type and heat transport medium, nuclear steam generators of the helix type are now in operation, installlation, fabrication or in the project phase. As a rule, projects are based on the extrapolation of existing technologies. In the particlular case of steam generators for HTGR power stations, however, existing experience is with steam generators of up to about 2 m diameter whereas several projects involve units more than twice as large. For this reason it was felt that a fabrication experiment was necessary in order to verify the feasibility of modern steam generator designs. A test rig was erected in the SULZER steam generator shops at Mantes, France, and skilled personnel and conventional production tools were employed in conducting experiments relating to the coiling, handling and threading of large helices. (Auth.)

Microbial analysis of the buffer/container experiment at AECL's underground research laboratory

International Nuclear Information System (INIS)

Stroes-Gascoyne, S.

1996-07-01

The Buffer/Container Experiment (BCE) was carried out at AECL's Underground Research Laboratory (URL) for 2.5 years to examine the in situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived the conditions (i.e., compaction, heat and desiccation) in the BCE and to determine which group(s) of microorganisms would be dominant in such a simulated vault environment. Such knowledge will be very useful in assessing the potential effects of microbial activity on the concept for deep disposal of Canada's nuclear fuel waste, proposed by AECL. 46 refs., 31 tabs., 35 figs

Capillary-Driven Heat Transfer Experiment: Keeping It Cool in Space

Science.gov (United States)

Lekan, Jack F.; Allen, Jeffrey S.

1998-01-01

Capillary-pumped loops (CPL's) are devices that are used to transport heat from one location to another--specifically to transfer heat away from something. In low-gravity applications, such as satellites (and possibly the International Space Station), CPL's are used to transfer heat from electrical devices to space radiators. This is accomplished by evaporating one liquid surface on the hot side of the CPL and condensing the vapor produced onto another liquid surface on the cold side. Capillary action, the phenomenon that causes paper towels to absorb spilled liquids, is used to "pump" the liquid back to the evaporating liquid surface (hot side) to complete the "loop." CPL's require no power to operate and can transfer heat over distances as large as 30 ft or more. Their reliance upon evaporation and condensation to transfer heat makes them much more economical in terms of weight than conventional heat transfer systems. Unfortunately, they have proven to be unreliable in space operations, and the explanation for this unreliability has been elusive. The Capillary-Driven Heat Transfer (CHT) experiment is investigating the fundamental fluid physics phenomena thought to be responsible for the failure of CPL's in low-gravity operations. If the failure mechanism can be identified, then appropriate design modifications can be developed to make capillary phase-change heat-transport devices a more viable option in space applications. CHT was conducted onboard the Space Shuttle Columbia during the first Microgravity Science Laboratory (MSL-1) mission, STS-94, which flew from July 1 to 17, 1997. The CHT glovebox investigation, which was conceived by Dr. Kevin Hallinan and Jeffrey Allen of the University of Dayton, focused on studying the dynamics associated with the heating and cooling at the evaporating meniscus within a capillary phase-change device in a low-gravity environment. The CHT experimental hardware was designed by a small team of engineers from Aerospace Design

First results of out-of-pile experiments concerning cooling phenomena of molten layers with internal heat sources

International Nuclear Information System (INIS)

Fieg, G.

1977-01-01

After severe hypothetical reactor accidents, large amounts of molten core material with internal heat generation may appear. It must be guaranteed that these materials can be kept within the containment. To clarify this situation, the knowledge of heat transport from liquid layers with internal heat generation is needed. First experimental results on heat transport from internally heated horizontal fluid layers are presented. The experiments have been performed in a smooth horizontal vessel with the base of 15 x 15 cm 2 . The Joule-heated liquid layer (depth L = 1 cm - 3.5 cm) is enclosed between two isothermal horizontal walls. They are polished fore parts of heat exchangers. The temperatures of the walls were held constant with thermostatically controlled water circulating through the heat exchangers. Horizontal heat fluxes were depressed by appropriate insulation of the side walls. The total heat transport to the upper and lower boundaries has been measured by the mass transport through the heat exchangers and the temperature rise of the cooling water

Development of Latent Heat Storage Phase Change Material Containing Plaster

Directory of Open Access Journals (Sweden)

Diana BAJARE

2016-05-01

Full Text Available This paper reviews the development of latent heat storage Phase Change Material (PCM containing plaster as in passive application. Due to the phase change, these materials can store higher amounts of thermal energy than traditional building materials and can be used to add thermal inertia to lightweight constructions. It was shown that the use of PCMs have advantages stabilizing the room temperature variations during summer days, provided sufficient night ventilation is allowed. Another advantage of PCM usage is stabilized indoor temperature on the heating season. The goal of this study is to develop cement and lime based plaster containing microencapsulated PCM. The plaster is expected to be used for passive indoor applications and enhance the thermal properties of building envelope. The plaster was investigated under Scanning Electron Microscope and the mechanical, physical and thermal properties of created plaster samples were determined.

Installation method for the steel container and vessel of the nuclear heating reactor

International Nuclear Information System (INIS)

Chen Liying; Guo Jilin; Liu Wei

2000-01-01

The Nuclear Heating Reactor (NHR) has the advantages of inherent safety and better economics, integrated arrangement, full power natural circulation and dual vessel structure. However, the large thin container presents a new and difficult problem. The characteristics of the dual vessel installation method are analyzed with system engineering theory. Since there is no foreign or domestic experience, a new method was developed for the dual vessel installation for the 5 MW NHR. The result shows that the installation method is safe and reliable. The research on the dual vessel installation method has important significance for the design, manufacture and installation of the NHR dual vessel, as well as the industrialization and standardization of the NHR

ICRF heating and current drive experiments on TFTR

International Nuclear Information System (INIS)

Rogers, J.H.; Hosea, J.C.; Phillips, C.K.

1996-01-01

Recent experiments in the Ion Cyclotron Range of Frequencies (ICRF) at TFTR have focused on the RF physics relevant to advanced tokamak D-T reactors. Experiments performed either tested confinement in reactor relevant plasmas or tested specific ICRF heating scenarios under consideration for reactors. H-minority heating was used to supply identical heating sources for matched D-T and D only L-mode plasmas to determine the species scaling for energy confinement. Second harmonic tritium heating was performed with only thermal tritium ions in an L-mode target plasma, verifying a possible start-up scenario for the International Thermonuclear Experimental Reactor (ITER). Direct electron heating in Enhanced Reverse Shear (ERS) plasmas has been found to delay the back transition out of the ERS state. D-T mode conversion of the fast magnetosonic wave to an Ion Berstein Wave (IBW) for off-axis heating and current drive has been successfully demonstrated for the first time. Parasitic Li 7 cyclotron damping limited the fraction of the power going to the electrons to less than 30%. Similar parasitic damping by Be 9 could be problematic in ITER. Doppler shifted fundamental resonance heating of beam ions and alpha particles has also been observed

Simulation of the ACE L2 and ACE L5 MCCI experiment under dry surface conditions with ASTEC MEDICIS using an effective heat transfer model

Energy Technology Data Exchange (ETDEWEB)

Agethen, Kathrin; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

2013-07-01

In a postulated severe accident the loss of cooling can lead to a melting of the core and to a failure of the vessel. The molten core material discharges to the containment cavity and interacts with the concrete basemat. The heat up of the concrete leads to the release of sparing gases (H{sub 2}, CO{sub 2}, SiO), which stir the pool und causes chemical reactions. Especially the metals (Zr, Fe, Ni, Cr) in the corium are oxidized und the exothermic energy is released to the melt, which raises the melt temperature further. The release of combustible gases (H{sub 2}, CO) and fission products to the containment atmosphere occurs as a result. In the long time (>10 h) containment failure and basemat penetration may occur, which can lead to fission product release to the environment. For further development and validation, simulations of experiments in which molten core concrete interaction (MCCI) is investigated, are necessary. In this work the new available effective heat transfer model in MEDICIS is used to calculate experiments of the ACE program, in which generic corium material is heated up and interacts with the concrete basemat. Here, especially the ACE L2 experiment with siliceous concrete and the ACE L5 experiment with limestone common sand (LCS) concrete will be presented. These experiments enable to analyze the heat transfer from the interior of the melt to the upper surface under dry conditions. Secondary the modeling in ASTEC version 2.p2 with the effective heat transfer module in MEDICIS is described. Results of MEDICIS simulations will be discussed by means of phenomena like ablation behavior and erosions depth, layer temperature and surface heat loss. Finally the issue of an effective heat transfer coefficient for the surface under dry conditions without top flooding is figured out. (orig.)

Full-scale and time-scale heating experiments at Stripa: preliminary results

International Nuclear Information System (INIS)

Cook, N.G.W.; Hood, Michael; California Univ., Berkeley

1978-01-01

Two full-scale heating experiments and a time-scale heating experiment have recently been started in granite 340 meters below surface. The purpose of the full-scale heating experiments is to assess the near-field effects of thermal loading for the design of an underground repository of nuclear wastes. That of the time-scale heating experiments is to obtain field data of the interaction between heaters and its effect on the rock mass during a period of about two years, which corresponds to about twenty years of full-scale operation. Geological features of the rock around each experiment have been mapped carefully, and temperatures, stresses and displacements induced in the rock by heating have been calculated in advance of the experiments. Some 800 different measurements are recorded at frequent intervals by a computer system situated underground. These data can be compared at any time with predictions made earlier on video display units underground

The buffer/container experiment: results, synthesis, issues

International Nuclear Information System (INIS)

Graham, J.; Chandler, N.A.; Dixon, D.A.; Roach, P.J.; To, T.; Wan, A.W.L.

1997-12-01

A large in-ground experiment has examined how heat affects the performance of the dense sand bentonite 'buffer' that has been proposed for use in the Canadian Nuclear Fuel Waste Management Program. The experiment was performed by Atomic Energy of Canada Limited at its Underground Research Laboratory, Lac du Bonnet, Manitoba between 1991 and 1994. The experiment placed a full-size heater representing a container of nuclear fuel waste in a 1.24-m diameter borehole filled with buffer below the floor of a room excavated at 240-m depth in granitic rock of the Canadian Shield. The buffer and surrounding rock were extensively instrumented for temperatures, total pressures, water pressures, suctions, and rock displacements. Power was provided to the heater for almost 900 days. The experiment showed that good rock conditions can be pre-selected, a borehole can be drilled, and buffer can be placed at controlled densities and water contents. The instrumentation generally worked well, and an extensive data base was successfully organized. Drying was observed in buffer close to the heater. This caused some desiccation cracking. However the cracks only extended approximately one third of the distance to the buffer-rock interface and did not form an advective pathway. Following sampling at the time of decommissioning, cracked samples of buffer were transported to the laboratory and given access to water. The hydraulic conductivities and swelling pressures of these resaturated samples were very similar to those of uncracked buffer. A good balance was achieved between the mass of water flowing into the experiment from the surrounding rock and the increased mass of water in the buffer. A good understanding was developed of the relationships between suctions, water contents, and total pressures in buffer near the buffer-rock interface. Comparisons between measurements and predictions of measured parameters show that a good understanding has been developed of the processes operating

The buffer/container experiment: results, synthesis, issues

Energy Technology Data Exchange (ETDEWEB)

Graham, J. [Univ. of Manitoba, Dept. of Civil Engineering, Winnipeg, MB (Canada); Chandler, N.A.; Dixon, D.A.; Roach, P.J.; To, T.; Wan, A.W.L

1997-12-01

A large in-ground experiment has examined how heat affects the performance of the dense sand bentonite 'buffer' that has been proposed for use in the Canadian Nuclear Fuel Waste Management Program. The experiment was performed by Atomic Energy of Canada Limited at its Underground Research Laboratory, Lac du Bonnet, Manitoba between 1991 and 1994. The experiment placed a full-size heater representing a container of nuclear fuel waste in a 1.24-m diameter borehole filled with buffer below the floor of a room excavated at 240-m depth in granitic rock of the Canadian Shield. The buffer and surrounding rock were extensively instrumented for temperatures, total pressures, water pressures, suctions, and rock displacements. Power was provided to the heater for almost 900 days. The experiment showed that good rock conditions can be pre-selected, a borehole can be drilled, and buffer can be placed at controlled densities and water contents. The instrumentation generally worked well, and an extensive data base was successfully organized. Drying was observed in buffer close to the heater. This caused some desiccation cracking. However the cracks only extended approximately one third of the distance to the buffer-rock interface and did not form an advective pathway. Following sampling at the time of decommissioning, cracked samples of buffer were transported to the laboratory and given access to water. The hydraulic conductivities and swelling pressures of these resaturated samples were very similar to those of uncracked buffer. A good balance was achieved between the mass of water flowing into the experiment from the surrounding rock and the increased mass of water in the buffer. A good understanding was developed of the relationships between suctions, water contents, and total pressures in buffer near the buffer-rock interface. Comparisons between measurements and predictions of measured parameters show that a good understanding has been developed of the processes

Effect of heating oils and fats in containers of different materials on their trans fatty acid content.

Science.gov (United States)

Kala, A L Amrutha; Joshi, Vishal; Gurudutt, K N

2012-08-30

The nature of the container material and temperature employed for deep-frying can have an influence on the development of trans fatty acids (TFAs) in the fat used. The present study was undertaken to determine the effect of heating vegetable oils and partially hydrogenated vegetable fats with different initial TFA content in stainless steel, Hindalium (an aluminium alloy), cast iron and glass containers. Ground nut oil (oil 1), refined, bleached and deodorised (RBD) palmolein (oil 2) and two partially hydrogenated vegetable oils with low (fat 1) and high (fat 2) TFA content were uniformly heated at 175-185 °C over a period of 12 h. An increase in TFA content to 20 g kg⁻¹ was observed in oil 2 in the cast iron container, while a decrease in TFA content of 20-30 g kg⁻¹ was observed in fat 2 in all containers. The heating process of fats and oils also led to an increase in Butyro refractometer reading and colour values. This study showed that the TFA 18:1t content of oil 1, oil 2 and fat 1 increased with repeated or prolonged heating. The cast iron container showed the highest increase in TFA 18:1t for RBD palmolein (oil 2). The amount of linoleic acid trans isomers formed in the heating process was negligible. Fat 2 with high initial TFA content showed a decrease in TFA 18:1 and 18:2 on heating in all containers. Oils heated in glass and stainless steel containers showed less TFA 18:1t formation. Copyright © 2012 Society of Chemical Industry.

Phase Change Materials-Assisted Heat Flux Reduction: Experiment and Numerical Analysis

Directory of Open Access Journals (Sweden)

Hussein J. Akeiber

2016-01-01

Full Text Available Phase change materials (PCM in the construction industry became attractive because of several interesting attributes, such as thermo-physical parameters, open air atmospheric condition usage, cost and the duty structure requirement. Thermal performance optimization of PCMs in terms of proficient storage of a large amount of heat or cold in a finite volume remains a challenging task. Implementation of PCMs in buildings to achieve thermal comfort for a specific climatic condition in Iraq is our main focus. From this standpoint, the present paper reports the experimental and numerical results on the lowering of heat flux inside a residential building using PCM, which is composed of oil (40% and wax (60%. This PCM (paraffin, being plentiful and cost-effective, is extracted locally from waste petroleum products in Iraq. Experiments are performed with two rooms of identical internal dimensions in the presence and absence of PCM. A two-dimensional numerical transient heat transfer model is developed and solved using the finite difference method. A relatively simple geometry is chosen to initially verify the numerical solution procedure by incorporating in the computer program two-dimensional elliptic flows. It is demonstrated that the heat flux inside the room containing PCM is remarkably lower than the one devoid of PCM.

Microwave heating device for internal heating convection experiments, applied to Earth's mantle dynamics.

Science.gov (United States)

Surducan, E; Surducan, V; Limare, A; Neamtu, C; Di Giuseppe, E

2014-12-01

We report the design, construction, and performances of a microwave (MW) heating device for laboratory experiments with non-contact, homogeneous internal heating. The device generates MW radiation at 2.47 GHz from a commercial magnetron supplied by a pulsed current inverter using proprietary, feedback based command and control hardware and software. Specially designed MW launchers direct the MW radiation into the sample through a MW homogenizer, devised to even the MW power distribution into the sample's volume. An adjustable MW circuit adapts the MW generator to the load (i.e., the sample) placed in the experiment chamber. Dedicated heatsinks maintain the MW circuits at constant temperature throughout the experiment. Openings for laser scanning for image acquisition with a CCD camera and for the cooling circuits are protected by special MW filters. The performances of the device are analyzed in terms of heating uniformity, long term output power stability, and load matching. The device is used for small scale experiments simulating Earth's mantle convection. The 30 × 30 × 5 cm(3) convection tank is filled with a water‑based viscous fluid. A uniform and constant temperature is maintained at the upper boundary by an aluminum heat exchanger and adiabatic conditions apply at the tank base. We characterize the geometry of the convective regime as well as its bulk thermal evolution by measuring the velocity field by Particle Image Velocimetry and the temperature field by using Thermochromic Liquid Crystals.

Abnormal excess heat observed during Mizuno-type experiments

International Nuclear Information System (INIS)

Fauvarque, Jean-Francois; Clauzon, Pierre Paul; Lalleve, Gerard Jean-Michel

2006-01-01

A simple calorimeter has been designed that works at constant temperature; that of boiling water. Heat Losses can be estimated accurately with an ohmic heater. As expected, losses are independent of the electric power input to the heater and the amount of evaporated water is linearly dependant on the power input. The device has been used to determine the heating power of a plasma electrolysis (the Ohmori-Mizuno experiment). We confirm that in this experiment, the heat output from electrolysis is greater than the electrical power input. The excess energy increases as the electrolysis voltage is increased from 200 up to 350 V (400 V input). The excess energy may be as high as 120 W. (author)

Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen

2005-08-29

This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

International Nuclear Information System (INIS)

Wright, A.L.; Kress, T.S.; Smith, A.M.

1980-01-01

This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

Comparison of heat stability of goat milk subjected to ultra-high temperature and in-container sterilization.

Science.gov (United States)

Chen, B Y; Grandison, A S; Lewis, M J

2012-03-01

Goat milk with and without stabilizing salt was subjected to in-container and UHT sterilization. Heat stability was assessed by measuring the amount of sediment in the milk. Without stabilizing salts, goat milk usually produced less sediment when subjected to in-container sterilization compared with UHT processing. Addition of stabilizing salts up to 12.8mM resulted in a progressive increase in sediment for in-container sterilization. In contrast, adding stabilizing salts at 6.4mM initially reduced sediment formation in UHT-treated milk but addition of stabilizing salts at 12.8mM increased sediment formation. Adding stabilizing salts to goat milk increased pH, decreased ionic calcium, and increased ethanol stability. Adding up to 2mM calcium chloride increased sediment formation more after UHT treatment than after in-container sterilization. These results suggest that no single mechanism or set of reactions causes milk to produce sediment during heating and that the favored pathway is different for UHT and in-container sterilization processes. Poor heat stability could be induced both by increasing ionic calcium and by decreasing it. Ethanol stability is not a good indicator of heat stability for in-container sterilization, but it may be for UHT sterilization, if milk does not enter the region of poor heat stability found at low concentrations of ionic calcium. Copyright © 2012 American Dairy Science Association. Published by Elsevier Inc. All rights reserved.

Heat, mass, and momentum transport model for hydrogen diffusion flames in nuclear reactor containments

International Nuclear Information System (INIS)

Travis, J.R.

1985-01-01

It is now possible to analyze the time-dependent, fully three-dimensional behavior of hydrogen diffusion flames in nuclear reactor containments. This analysis involves coupling the full Navier-Stokes equations with multi-species transport to the global chemical kinetics of hydrogen combustion. A transport equation for the subgrid scale turbulent kinetic energy density is solved to produce the time and space dependent turbulent transport coefficients. The heat transfer coefficient governing the exchange of heat between fluid computational cells adjacent to wall cells is calculated by a modified Reynolds analogy formulation. The analysis of a MARK-III containment indicates very complex flow patterns that greatly influence fluid and wall temperatures and heat fluxes. 18 refs., 24 figs

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

International Nuclear Information System (INIS)

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

1991-08-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-08-01

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

Results of an experiment in a Zion-like geometry to investigate the effect of water on the containment basement floor on direct containment heating (DCH) in the Surtsey Test Facility: The IET-4 test

International Nuclear Information System (INIS)

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

1992-09-01

This document discusses the fourth experiment of the Integral Effects Test (IET-4) series which was conducted to investigate the effects of high pressure melt ejection on direct containment heating. Scale models (1:10) of the Zion reactor pressure vessel (RPV), cavity, instrument tunnel, and subcompartment structures were constructed in the Surtsey Test Facility at Sandia National Laboratories. ne RPV was modeled with a melt generator that consisted of a steel pressure barrier, a cast MgO crucible, and a thin steel inner liner. The melt generator/crucible had a hemispherical bottom head containing a graphite limitor plate with a 3.5-cm exit hole to simulate the ablated hole in the RPV bottom head that would be tonned by tube ejection in a severe nuclear power plant accident. The reactor cavity model contained 3.48 kg of water with a depth of 0.9 cm that corresponded to condensate levels in the Zion plant. A 43-kg initial charge of iron oxide/aluminum/chromium thermite was used to simulate corium debris on the bottom head of the RPV. Molten thermite was ejected into the scaled reactor cavity by 6.7 MPa steam. IET-4 replicated the third experiment in the IET series (IET-3), except the Surtsey vessel contained slightly more preexisting oxygen (9.6 mol.% vs. 9.0 mol.%), and water was placed on the basement floor inside the crane wall. The cavity pressure measurements showed that a small steam explosion occurred in the cavity at about the same time as the steam explosion in IET-1. The oxygen in the Surtsey vessel in IET-4 resulted in a vigorous hydrogen bum, which caused a significant increase in the peak pressure, 262 kPa compared to 98 kPa in the IET-1 test. EET-3, with similar pre-existing oxygen concentrations, also had a large peak pressure of 246 kPa

A study on experiment and numerical simulation of heat exchanger in heating furnace

Directory of Open Access Journals (Sweden)

Z. C. Lv

2018-01-01

Full Text Available In this paper, air preheater is used the research object and its heat transfer law is studied by experiment and numerical simulation. The experimental data showed that with the increases of inlet air velocity, the comprehensive heat transfer coefficient and heat transfer efficiency increase, but the temperature efficiency decreases and the resistance loss on the air side increases. The numerical simulation results showed that the larger the diameter of the tube, the better the heat transfer effect. When horizontal spacing in the range of 290 - 305 mm and longitudinal spacing is 70 - 90 mm, the heat transfer effect is best. The optimized heat exchanger structure is that diameter is 60 mm, horizontal spacing is 300 mm, longitudinal spacing is 90 mm. As the inlet air flow rate increases, the heat transfer efficiency increases, but the temperature efficiency decreases and the resistance loss on the air side increases.

Coupling heat and chemical tracer experiments for estimating heat transfer parameters in shallow alluvial aquifers.

Science.gov (United States)

Wildemeersch, S; Jamin, P; Orban, P; Hermans, T; Klepikova, M; Nguyen, F; Brouyère, S; Dassargues, A

2014-11-15

Geothermal energy systems, closed or open, are increasingly considered for heating and/or cooling buildings. The efficiency of such systems depends on the thermal properties of the subsurface. Therefore, feasibility and impact studies performed prior to their installation should include a field characterization of thermal properties and a heat transfer model using parameter values measured in situ. However, there is a lack of in situ experiments and methodology for performing such a field characterization, especially for open systems. This study presents an in situ experiment designed for estimating heat transfer parameters in shallow alluvial aquifers with focus on the specific heat capacity. This experiment consists in simultaneously injecting hot water and a chemical tracer into the aquifer and monitoring the evolution of groundwater temperature and concentration in the recovery well (and possibly in other piezometers located down gradient). Temperature and concentrations are then used for estimating the specific heat capacity. The first method for estimating this parameter is based on a modeling in series of the chemical tracer and temperature breakthrough curves at the recovery well. The second method is based on an energy balance. The values of specific heat capacity estimated for both methods (2.30 and 2.54MJ/m(3)/K) for the experimental site in the alluvial aquifer of the Meuse River (Belgium) are almost identical and consistent with values found in the literature. Temperature breakthrough curves in other piezometers are not required for estimating the specific heat capacity. However, they highlight that heat transfer in the alluvial aquifer of the Meuse River is complex and contrasted with different dominant process depending on the depth leading to significant vertical heat exchange between upper and lower part of the aquifer. Furthermore, these temperature breakthrough curves could be included in the calibration of a complex heat transfer model for

Status of direct containment heating in CSNI member countries. Report of task group on ex-vessel thermal-hydraulics

International Nuclear Information System (INIS)

1989-03-01

The status of activities on direct containment heating in the light water reactor program in OECD/CSNI countries is presented. Experimental and analytical studies are reviewed. Approaches or measures are discussed for accident management in relation to direct containment heating. A discussion is given of common and diverging views among the countries based, in part, on response to a questionnaire. The key issues are discussed and recommendations are provided for future CSNI work on direct containment heating

Subseabed Disposal Program In-Situ Heat Transfer Experiment (ISHTE)

International Nuclear Information System (INIS)

Percival, C.M.

1983-05-01

A heat transfer experiment is being developed in support of the Subseabed Disposal Program. The primary objectives of this experiment are: to provide information on the in situ response of seabed sediment to localized heating; to provide an opportunity to evaluate theoretical models of the response and to observe any unanticipated phenomena which may occur; and to develop and demonstrate the technology necessary to perform waste isolation oriented experiments on the seafloor at depths up to 6000 m. As presently envisaged, the heat transfer experiment will be conducted at a location in the central North Pacific though it could be performed anywhere that the ocean bottom is of the type deemed suitable for the disposal of nuclear waste material. The experiment will be conducted of the seafloor from a recoverable space-frame platform at a depth of approximately 6000 m. A 400-W isotopic heat source will be implanted in the illite sediment and the subsequent response of the sediment to the induced thermal field evaluated. After remote initiation of the experiment, a permanent record of the data obtained will be recorded on board the platform, with selected information transmitted to a surface vessel by acoustic telemetry. The experiment will be operational for one year, after which the entire platform will be recovered. Current plans call for the deployment of the experiment in 1986. Specific activities which will be pursued during the course of the experiment include: measurement of the thermal field; determination of the effective thermal conductivity of the sediment; measurement of pore pressure; evaluation of radionuclide migration processes; pore water sampling; sediment chemistry studies; sediment shear strength measurements; and coring operations in the immediate vicinity of the experiment for postexperiment analysis

Post-accident heat removal ''information exchange''

International Nuclear Information System (INIS)

Plein, H.G.; Carlson, G.A.

1975-01-01

The in-core molten pool experiments are designed to produce a pool of fission heated temperature and flow patterns of such pools, and evaluate the barrier melt-through potential of the molten UO 2 . The first experiments, to be conducted this fiscal year in the Annular Core Pulse Reactor, will be uncomplicated and multiply-contained to prove containment design and to provide initial information on fission heated molten pool characteristics. Concurrent with the in-core experiments, high temperature ultrasonic techniques are being developed to measure UO 2 temperatures up to and above the melting point for use in later more definitive experiments scheduled for FY77

Mathematical modelling of heat absorption capacity of containment spray system in a 700 MWe PHWR

International Nuclear Information System (INIS)

Kota, Sampath Bharadwaj; Ali, Seik Mansoor; Balasubramaniyan, V.

2015-01-01

This paper presents a mathematical model for estimating the heat removal by containment spray system in the post Loss of Coolant Accident (LOCA) environment. The procedure involves firstly, the calculation of heat removal rates by droplets of spray dispersed in the air-steam mixture by an appropriate direct contact condensation model accounting for the presence of non-condensable gas (air). Parametric influence of droplet size, ambient pressure and temperature on heat flux is brought out. It was found that the heat flux is inversely proportional to the ambient pressure and diameter. A spray module was subsequently developed and incorporated into an in-house containment thermal hydraulics code. The pressure and temperature transients in a 700 MWe PHWR containment building following a Large Break LOCA was obtained using this code. The efficacy of the spray in condensing the steam is shown by comparing the transients with and without the operation of spray system. Parametric studies are also conducted with respect to droplet size and flow rate of water droplet spray. The details of the investigation are presented and discussed in this paper. (author)

Nuclear heat applications in Russia: Experience, status and prospects

International Nuclear Information System (INIS)

Mitenkov, F.M.; Kusmartsev, E.V.

1998-01-01

The extensive experience gained with nuclear district heating in Russia is described. Most of the WWER reactors in Russia are cogeneration plants. Steam is extracted through LP turbine bleeders and condensed in intermediate heat exchangers to hot water which is then supplied to DH grids. Also some small dedicated nuclear heating plants are operated. (author)

29 CFR 1915.54 - Welding, cutting and heating of hollow metal containers and structures not covered by § 1915.12.

Science.gov (United States)

2010-07-01

... 29 Labor 7 2010-07-01 2010-07-01 false Welding, cutting and heating of hollow metal containers and... STANDARDS FOR SHIPYARD EMPLOYMENT Welding, Cutting and Heating § 1915.54 Welding, cutting and heating of... which have contained flammable substances shall, before welding, cutting, or heating is undertaken on...

Theory of ionospheric heating experiments

International Nuclear Information System (INIS)

Cragin, B.L.

1975-01-01

A brief description of the F region ionospheric heating experiments is given including some historical notes and a brief summary of the observations. A theory for the phenomenon of ''artificial spread F'' is presented. The explanation is in terms of scattering by approximately field-aligned, large scale ionization density irregularities, which are produced by a thermal version of the stimulated Brillouin scattering instability in which the heating wave decays into another electromagnetic wave and an electrostatic wave of very low frequency. This thermal instability differs from conventional stimulated Brillouin scattering in that the low frequency wave is driven by differential heating in the interference pattern of the two electromagnetic waves, rather than by the usual ponderomotive force. Some aspects of the theory of the phenomenon of ''wide-band attenuation'' or ''anomalous absorption'' of a probing electromagnetic wave. Some general results from the theory of wave propagation in a random medium are used to derive equations describing the absorption of a probing electromagnetic wave due to scattering (by large scale irregularities) into new electromagnetic waves or (by small scale irregularities) into electron plasma oscillations

Heat transfer analysis of the waste-container sleeve/salt configuration

International Nuclear Information System (INIS)

Callahan, G.D.; Ratigan, J.L.; Russell, J.E.; Fossum, A.F.

1975-01-01

Prior to this investigation, the heat transport considered was only that of straight conduction. The waste container, air gap, and sleeve arrangement was considered to be a single, consistent, time-dependent, heat-generating unit in intimate contact with the salt. The conduction model does not accurately model the heat transfer mechanisms available. Thus radiation and combined radiation and convection must also be considered in the determination of the temperature field. As would be expected, the canister temperatures are higher for the case of radiation across the airgap than those that result from conduction when the canister is in intimate contact with the salt. For the radiation case, the canister temperatures rise rapidly to a temperature of approximately 1,140 0 F and maintain an almost steady state condition for one year whereafter the temperatures slowly decrease. The far field temperatures, near the pillar centerline, are essentially equivalent for all cases. As time proceeds, the far field temperatures of the conduction models are about 15% different

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

Energy Technology Data Exchange (ETDEWEB)

Li, Cheng, E-mail: 510395453@qq.com [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China); Li, Le [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Junming [Tsinghua University, Key Laboratory for Thermal Science and Power Engineering of Ministry of Education, Department of Thermal Engineering, Beijing 100084 (China); Zhang, Yajun [Tsinghua University, Institute of Nuclear and New Energy Technology, 100084 Beijing (China); Li, Zhihui [State Nuclear Power Technology Research & Development Center, 102209 Beijing (China)

2017-03-15

Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

Analysis of the passive heat removal enhancement for AP1000 containment due to the partially wetted coverage

International Nuclear Information System (INIS)

Li, Cheng; Li, Le; Li, Junming; Zhang, Yajun; Li, Zhihui

2017-01-01

Highlights: • Heat removal by steam condensation, thermal conduction and evaporation is the most important scheme for AP1000 PCCS. Traditionally, studies on containment wall condensation and evaporation have been widely made, while it lacks studies on the shell two-dimension (2-D) thermal conduction. Currently, based on the known heat and mass transfer correlations and the phenomenon from water wetted coverage test, the physical model for 2-D thermal conduction is given and numerical simulation is then made. By discussions, it forms the following highlights. • The partially wetted surface can enhance the whole heat transfer process (including inner condensation, wall thermal conduction and outside cooling) and the maximum enhancement factor can be as large as 63%. There is an enhancement peak at around dry strip fraction a = 90%. When L is less than 0.03 m, its influence on heat transfer is small and the enhancement is mainly affected by dry coverage. However, for larger L, both α and L contribute much to larger enhancement. • Location at the spring line is often used for safety analysis and the dry strip fraction there for AP1000 is mainly at 10%–80%. Accordingly, further analysis is made on L (0.03 < L < 0.3) and a fitting expression is given for α = 10%–80%. It could be used to improve the corresponding software and it could also be used for containment scaling-down criteria analysis. - Abstract: AP1000 containment uses the water film evaporation, coupled with containment inner condensation, to remove the core decay heat. However, water film cannot fully cover heat transfer surface and dry-wetted strips appear. As a result, heat transfer within the containment shell is a two-dimension thermal conduction. Current work numerically studied the AP1000 heat removal enhancement due to the partially wetted coverage phenomenon. It used the evaporation and condensation boundary conditions and Fluent software to calculate the local heat fluxes and their

Experiment on transient heat transfer in closed narrow channel

International Nuclear Information System (INIS)

Ochiai, Masaaki

1985-01-01

Heat transfer coefficients and transient pressures in closed narrow channels were obtained experimentally, in order to assess the gap heat transfer models in the computer code WTRLGD which were devised to analyze the internal pressure behavior of waterlogged fuel rods. Gap widths of channels are 0.1--0.5mm to simulate the gap region of waterlogged fuel rods, and test fluids are water (7--89.2 0 C) and Freon-113 (9.2 0 C). The results show that the heater temperature and the pressure measured in the experiments without the DNB occurrence are simulated fairly well by the calculational model of WTRLGD where the heat transfer in a closed narrow channel is evaluated with one-dimensional transient thermal conduction equation and Jens and Lottes' correlation for nucleate boiling. Consequently, it is also suggested that the above equations are available for evaluation of heat flux from fuel to internal water of waterlogged fuel rods. The film boiling heat transfer coefficient was in the same order of that evaluated by Bromley's correlation and the DNB heat flux was smaller than that obtained in quasi-steady experiments with ordinary systems, although the experimental data for them were not enough. (author)

Ion Bernstein wave heating experiments in HT-7 superconducting tokamak

International Nuclear Information System (INIS)

Zhao Yanping

2005-01-01

Ion Bernstein Wave (IBW) experiments have been carried out in recent years in the HT-7 superconducting Tokamak. The electron heating experiment has been concentrated on deuterium plasma with an injecting RF power up to 350 kw. The globe heating and localized heating can be seen clearly by controlling the ICRF resonance layer's position. On-axis and off-axis electron heating have been realized by properly setting the target plasma parameters. Experimental results show that the maximum increment in electron temperature has been more than 1 keV, the electron temperature profile has been modified by IBW under different plasma conditions, and both energy and particle confinement improvements have been obtained. (author)

Scaling options for integral experiments for molten salt fluid mechanics and heat transfer

International Nuclear Information System (INIS)

Philippe Bardet; Per F Peterson

2005-01-01

experiments can reproduce molten salt phenomena including natural, mixed, and forced convection fluid mechanics and heat transfer, and free surface fluid mechanics, with very small scaling distortion. Weber number controlled phenomena like bubble entrainment and droplet formation are also simulated with relatively low distortion. Both molten salts and mineral oils are transparent, permitting flow visualization to be used, and the oil index of refraction nearly matches acrylic. Oils have much lower wetting angles than molten salts with typical container materials, so significant distortion of wetting-angle related phenomena, such as capillary driven flows, can be expected. Overall, scaled experiments using light mineral oils have very attractive benefits for studying molten salt fluid mechanics and heat transfer. This paper discusses scaling in detail, and presents examples from fusion chamber fluid mechanics experiments. (authors)

Boiling Heat Transfer Coefficients of Nanofluids Containing Carbon Nanotubes up to Critical Heat Fluxes

International Nuclear Information System (INIS)

Park, Ki Jung; Lee, Yohan; Jung, Dong Soo; Shim, Sang Eun

2011-01-01

In this study, the nucleate pool boiling heat transfer coefficients (HTCs) and critical heat flux (CHF) for a smooth and square flat heater in a pool of pure water with and without carbon nanotubes (CNTs) dispersed at 60 .deg. C were measured. Tested aqueous nanofluids were prepared using CNTs with volume concentrations of 0.0001%, 0.001%, and 0.01%. The CNTs were dispersed by chemically treating them with an acid in the absence of any polymers. The results showed that the pool boiling HTCs of the nanofluids are higher than those of pure water in the entire nucleate boiling regime. The acid-treated CNTs led to the deposition of a small amount of CNTs on the surface, and the CNTs themselves acted as heat-transfer-enhancing particles, owing to their very high thermal conductivity. There was a significant increase in the CHF- up to 150%-when compared to that of pure water containing CNTs with a volume concentration of 0.001%. This is attributed to the change in surface characteristics due to the deposition of a very thin layer of CNTs on the surface. This layer delays nucleate boiling and causes a reduction in the size of the large vapor canopy around the CHF. This results in a significant increase in the CHF

Design of an experiment to measure fire exposure of packages aboard container cargo ships

International Nuclear Information System (INIS)

Koski, J.A.

1998-01-01

The test described in this paper is intended to measure the typical accident environment for a radioactive materials package in a fire abroad a container cargo ship. A stack of nice used standard cargo containers will be variously loaded with empty packages, simulated packages and combustible cargo and placed over a large hydrocarbon pool fire of one hour duration. Fire environments, both inside and outside the containers, typical of on-deck stowage will be measured as well as the potential for container-to-container fire spread. With the use of the inverse heat conduction calculations, the local heat transfer to the simulated packages can be estimated from thermocouple data. Data recorded will also provide information on fire durations in each container, fire intensity and container-to-container fire spread characteristics. (authors)

Plasma heating by relativistic electron beams: correlations between experiment and theory

International Nuclear Information System (INIS)

Thode, L.E.; Godfrey, B.B.

1975-01-01

The streaming instability is the primary heating mechanism in most, if not all, experiments in which the beam is injected into partially or fully ionized gas. In plasma heating experiments, the relativistic beam must traverse an anode foil before interacting with the plasma. The linear theory for such a scattered beam is discussed, including a criterion for the onset of the kinetic interaction. A nonlinear model of the two-stream instability for a scattered beam is developed. Using this model, data from ten experiments are unfolded to obtain the following correlations: (i) for a fixed anode foil, the dependence of the plasma heating on the beam-to-plasma density ratio is due to anode foil scattering, (ii) for a fixed beam-to-plasma density ratio, the predicted change in the magnitude of plasma heating as a function of the anode foil is in agreement with experiment, and (iii) the plasma heating tentatively appears to be proportional to the beam kinetic energy density and beam pulse length. For a fixed anode foil, theory also predicts that the energy deposition is improved by increasing the beam electron energy γmc 2 . Presently, no experiment has been performed to confirm this aspect of the theory

In situ heating experiments in hard rock: their objectives and design

International Nuclear Information System (INIS)

Cook, N.G.W.; Witherspoon, P.A.; California Univ., Berkeley

1978-01-01

Of the many alternatives that are being considered for the disposal of nuclear wastes, deep underground burial is favored. The wealth of experience concerning the design and construction of underground excavations does not include the unique effects of heating excavations by radioactive decay, nor the issue of long-term isolation. The effects of heating are important in establishing the feasibility of this method of disposal and are essential for the design of an underground repository. Near-field phenomena around individual canisters can be studied by full-scale experiments, using electrical heaters. The thermal diffusivity of rock is so low that information concerning the interaction between full-scale heaters and of the effects of heating a large volume of rock cannot be measured in full-scale experiments lasting less than a few decades. To overcome this difficulty, a time-scaled heating experiment has been developed in which a reduction in linear scale is accompanied by an acceleration of the time scale to the second power. In this experiment, the linear scale is about a third, so that the time scale is about ten fold

In situ heating experiments in hard rock: their objectives and design

International Nuclear Information System (INIS)

Cook, N.G.W.; Witherspoon, P.A.

1978-01-01

Of the many alternatives that are being considered for the disposal of nuclear wastes, deep underground burial is favored. The wealth of experience concerning the design and construction of underground excavations does not include the unique effects of heating excavations by radioactive decay, nor the issue of long-term isolation. The effects of heating are important in establishing the feasibility of this method of disposal, and are essential for the design of an underground repository. Near-field phenomena around individual canisters can be studied by full-scale experiments, using electrical heaters. The thermal diffusivity of rock is so low that information concerning the interaction between full-scale heaters and of the effects of heating a large volume of rock cannot be measured in full-scale experiments lasting less than a few decades. To overcome this difficulty, a time-scaled heating experiment has been developed in which a reduction in linear scale is accompanied by an acceleration of the time scale to the second power. In this experiment, the linear scale is about a third, so that the time scale is about ten fold

Design experiments for a vented containment

International Nuclear Information System (INIS)

Hesboel, R.

1985-01-01

A filtered containment venting system, operable late in 1985, is currently under installation at the Barsebaeck twin nuclear power station in Sweden. The filter unit, which communicates with the containments of both reactor units, but is separated from them by rupture discs, consists of a concrete bed, 40 m high and 20 m in diameter, filled with gravel of grain size 25-35 mm. The performance of the gravel bed under such accident conditions which might lead to an activation of this safeguard system has been the subject for investigation within the FILTRA project. These investigations have shown that the gravel bed acts as: an expansion volume for decreasing gas pressure and increasing gas residence time, a heat sink for condensing steam, an excellent filter medium for removing aerosols and elemental iodine, and a sump volume for collecting radioactive condensate. The results from iodine retention studies in gravel beds are mainly considered

Microwave heating device for internal heating convection experiments, applied to Earth's mantle dynamics

Energy Technology Data Exchange (ETDEWEB)

Surducan, E.; Surducan, V.; Neamtu, C., E-mail: camelia.neamtu@itim-cj.ro [National Institute for Research and Development of Isotopic and Molecular Technologies (INCDTIM), 67-103 Donat St., 400293, Cluj‑Napoca (Romania); Limare, A.; Di Giuseppe, E. [Institut de Physique du Globe de Paris (IPGP), Univ. Paris Diderot, UMR CNRS 7154, 1 rue Jussieu, 75005, Paris (France)

2014-12-15

We report the design, construction, and performances of a microwave (MW) heating device for laboratory experiments with non-contact, homogeneous internal heating. The device generates MW radiation at 2.47 GHz from a commercial magnetron supplied by a pulsed current inverter using proprietary, feedback based command and control hardware and software. Specially designed MW launchers direct the MW radiation into the sample through a MW homogenizer, devised to even the MW power distribution into the sample's volume. An adjustable MW circuit adapts the MW generator to the load (i.e., the sample) placed in the experiment chamber. Dedicated heatsinks maintain the MW circuits at constant temperature throughout the experiment. Openings for laser scanning for image acquisition with a CCD camera and for the cooling circuits are protected by special MW filters. The performances of the device are analyzed in terms of heating uniformity, long term output power stability, and load matching. The device is used for small scale experiments simulating Earth's mantle convection. The 30 × 30 × 5 cm{sup 3} convection tank is filled with a water‑based viscous fluid. A uniform and constant temperature is maintained at the upper boundary by an aluminum heat exchanger and adiabatic conditions apply at the tank base. We characterize the geometry of the convective regime as well as its bulk thermal evolution by measuring the velocity field by Particle Image Velocimetry and the temperature field by using Thermochromic Liquid Crystals.

High temperature heat pipe experiments in low earth orbit

International Nuclear Information System (INIS)

Woloshun, K.; Merrigan, M.A.; Sena, J.T.; Critchley, E.

1993-01-01

Although high temperature, liquid metal heat pipe radiators have become a standard component on most high power space power system designs, there is no experimental data on the operation of these heat pipes in a zero gravity or micro-gravity environment. Experiments to benchmark the transient and steady state performance of prototypical heat pipe space radiator elements are in preparation for testing in low earth orbit. It is anticipated that these heat pipes will be tested aborad the Space Shuttle in 1995. Three heat pipes will be tested in a cargo bay Get Away Special (GAS) canister. The heat pipes are SST/potassium, each with a different wick structure; homogeneous, arterial, and annular gap, the heat pipes have been designed, fabricated, and ground tested. In this paper, the heat pipe designs are specified, and transient and steady-state ground test data are presented

Ion heating in minority ICRH experiments on JET

International Nuclear Information System (INIS)

Start, D.F.H.; Bhatnagar, V.; Bures, M.

1991-06-01

Bulk ion heating by high power H-minority ICRH has been demonstrated in JET during both pellet enhanced performance H-mode experiments (PEP + H - mode) and in density limit studies. In the PEP + H - mode plasmas the electron and ion temperatures both reached 10 keV at an electron density of 7 x 10 19 /m 3 . According to Fokker-Planck calculations the power from the minority was transfered almost equally to the electrons and majority ions as a result of both the high electron density, n e , and the high minority density, n h , (n h /n e ≅ 0.15). For the first time with ICRH on JET a central ion temperature greater than the central electron temperature was achieved. In the density limit experiments which involved strong gas puffing into limiter discharges, there was strong evidence of a transfer from electron heating to ion heating as the electron density was ramped up to 8 x 10 19 /m 3 . (Author)

Problems of heat transfer within the containing vessel of high performance LMFBR spent fuel shipping casks

International Nuclear Information System (INIS)

Pope, R.B.; Gartling, D.K.; Schimmel, W.P. Jr.; Larson, D.W.

1976-01-01

A preliminary assessment of heat transfer problems internal to a LMFBR spent fuel shipping cask is reported. The assessment is based upon previous results obtained in full-scale, electrically heated mockups of an LMFBR assembly located in a containing pipe, and also upon analytical and empirical studies presented in this paper. It is shown that a liquid coolant will be required to adequately distribute the decay heat of short-cooled assemblies from the fuel region to the containing cask structure. Liquid sodium apparently provides the best heat transfer, and sufficient data are available to adequately model the heat transfer processes involved. Dowtherm A is the most efficient organic evaluated to date and presented in the open literature. Since the organic materials have high Prandtl and usually high Rayleigh numbers, natural convection is the predominant mode of heat transfer. It is shown that a more comprehensive understanding of the convective processes will be required before heat transfer with an organic coolant can be adequately modeled. However, in view of systems considerations, Dowtherm A should be further considered as an alternative to sodium for use as a LMFBR spent fuel shipping cask coolant

The Cryogenic Test Bed experiments: Cryogenic heat pipe flight experiment CRYOHP (STS-53). Cryogenic two phase flight experiment CRYOTP (STS-62). Cryogenic flexible diode flight experiment CRYOFD

Science.gov (United States)

Thienel, Lee; Stouffer, Chuck

1995-09-01

This paper presents an overview of the Cryogenic Test Bed (CTB) experiments including experiment results, integration techniques used, and lessons learned during integration, test and flight phases of the Cryogenic Heat Pipe Flight Experiment (STS-53) and the Cryogenic Two Phase Flight Experiment (OAST-2, STS-62). We will also discuss the Cryogenic Flexible Diode Heat Pipe (CRYOFD) experiment which will fly in the 1996/97 time frame and the fourth flight of the CTB which will fly in the 1997/98 time frame. The two missions tested two oxygen axially grooved heat pipes, a nitrogen fibrous wick heat pipe and a 2-methylpentane phase change material thermal storage unit. Techniques were found for solving problems with vibration from the cryo-collers transmitted through the compressors and the cold heads, and mounting the heat pipe without introducing parasitic heat leaks. A thermally conductive interface material was selected that would meet the requirements and perform over the temperature range of 55 to 300 K. Problems are discussed with the bi-metallic thermostats used for heater circuit protection and the S-Glass suspension straps originally used to secure the BETSU PCM in the CRYOTP mission. Flight results will be compared to 1-g test results and differences will be discussed.

Evaporation and condensation heat transfer in a suppression chamber of the water wall type passive containment cooling system

International Nuclear Information System (INIS)

Fujii, Tadashi; Kataoka, Yoshiyuki; Murase, Michio

1996-01-01

To evaluate the system pressure response of a water wall type containment cooling system, which is one of the passive safety systems, the evaporation and condensation behaviors in a suppression chamber have been experimentally examined. In the system, the suppression pool water evaporates from the pool surface, passing into the wetwell due to pool temperature rise, while steam in the wetwell condenses on the steel containment vessel wall due to the heat release through the wall. The wetwell is a gas phase region in the suppression chamber and its pressure, which is expressed as the sum of the noncondensable gas pressure and saturated steam pressure, is strongly affected by the evaporation heat transfer from the suppression pool surface and condensation heat transfer on the containment vessel wall. Based on the measured temperature profiles near the heat transfer surface and the wetwell pressure using two apparatuses, evaporation and condensation heat transfer coefficients were evaluated. The following results were obtained. (1) Both heat transfer coefficients increased as the ratio of the steam partial pressure to the total pressure increased. (2) Comparison of the results from two types of test apparatuses confirmed that the size of the heat transfer surface did not affect the heat transfer characteristics within these tests. (3) The heat transfer coefficients were expressed by the ratio of the steam to noncondensable gas logarithmic mean concentration, which considered the steam and gas concentration gradient from the heat transfer surface to the wetwell bulk. (author)

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

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

1991-01-01

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

Recent DIII-D high power heating and current drive experiments

International Nuclear Information System (INIS)

Simonen, T.C.; Jackson, G.L.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.; Lazarus, E.A.

1994-02-01

This paper describes recent DIII-D high power heating and current drive experiments. Describes are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal beta

The design of a heat transfer liquid metal MHD experiment for ALEX [Argonne Liquid-Metal Experiment

International Nuclear Information System (INIS)

Picologlou, B.F.; Reed, C.B.; Hua, T.Q.; Lavine, A.S.

1988-01-01

An experiment to study heat transfer in liquid metal MHD flow, under conditions relevant to coolant channels for tokamak first wall and high heat flux devices, is described. The experimental configuration is a rectangular duct in a transverse magnetic field, heated on one wall parallel to the field. The specific objective of the experiment is to resolve important issues related to the presence and heat transfer characteristics of wall jets and flow instabilities in MHD flows in rectangular duct with electrically conducting walls. Available analytical tools for MHD thermal hydraulics have been used in the design of the test article and its instrumentation. Proposed tests will cover a wide range of Peclet and Hartmann numbers and interaction parameters. 14 refs., 3 figs., 1 tab

Recent DIII-D high power heating and current drive experiments

International Nuclear Information System (INIS)

Simonen, T.C.; Jackson, G.L.; Lazarus, E.A.; Mahdavi, M.A.; Petrie, T.W.; Politzer, P.A.; Taylor, T.S.

1995-01-01

This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal β. ((orig.))

Recent DIII-D high power heating and current drive experiments

Energy Technology Data Exchange (ETDEWEB)

Simonen, T.C. [General Atomics, San Diego, CA (United States); Jackson, G.L. [General Atomics, San Diego, CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Politzer, P.A. [General Atomics, San Diego, CA (United States); Taylor, T.S. [General Atomics, San Diego, CA (United States); DIII-D Team

1995-01-01

This paper describes recent DIII-D high power heating and current drive experiments. Described are experiments with improved wall conditioning, divertor particle pumping, radiative divertor experiments, studies of plasma shape and high poloidal {beta}. ((orig.)).

Docker Containers for Deep Learning Experiments

OpenAIRE

Gerke, Paul K.

2017-01-01

Deep learning is a powerful tool to solve problems in the area of image analysis. The dominant compute platform for deep learning is Nvidia’s proprietary CUDA, which can only be used together with Nvidia graphics cards. The nivida-docker project allows exposing Nvidia graphics cards to docker containers and thus makes it possible to run deep learning experiments in docker containers.In our department, we use deep learning to solve problems in the area of medical image analysis and use docker ...

Capillary Pumped Heat Transfer (CHT) Experiment

Science.gov (United States)

Hallinan, Kevin P.; Allen, J. S.

1998-01-01

The operation of Capillary Pumped Loops (CPL's) in low gravity has generally been unable to match ground-based performance. The reason for this poorer performance has been elusive. In order to investigate the behavior of a CPL in low-gravity, an idealized, glass CPL experiment was constructed. This experiment, known as the Capillary-driven Heat Transfer (CHT) experiment, was flown on board the Space Shuttle Columbia in July 1997 during the Microgravity Science Laboratory mission. During the conduct of the CHT experiment an unexpected failure mode was observed. This failure mode was a result of liquid collecting and then eventually bridging the vapor return line. With the vapor return line blocked, the condensate was unable to return to the evaporator and dry-out subsequently followed. The mechanism for this collection and bridging has been associated with long wavelength instabilities of the liquid film forming in the vapor return line. Analysis has shown that vapor line blockage in present generation CPL devices is inevitable. Additionally, previous low-gravity CPL tests have reported the presence of relatively low frequency pressure oscillations during erratic system performance. Analysis reveals that these pressure oscillations are in part a result of long wavelength instabilities present in the evaporator pores, which likewise lead to liquid bridging and vapor entrapment in the porous media. Subsequent evaporation to the trapped vapor increases the vapor pressure. Eventually the vapor pressure causes ejection of the bridged liquid. Recoil stresses depress the meniscus, the vapor pressure rapidly increases, and the heated surface cools. The process then repeats with regularity.

Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

International Nuclear Information System (INIS)

Langhans, J.

1991-01-01

Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

Behaviour of the Callovo-Oxfordian clay around a converging heated borehole: thermal free wall experiment

International Nuclear Information System (INIS)

Garitte, B.; Gens, A.; Vaunat, J.; Armand, G.; Conil, N.

2012-01-01

Document available in extended abstract form only. ANDRA has launched several heating experiments in the Meuse-Haute Marne Underground Laboratory (e.g. TER and TED). In these experiments, the heater-rock contact was ensured by a metal tubing that prevented any convergence of the heating borehole. The Thermal Free Wall experiment was run by ANDRA to investigate whether the rock behaviour around an un-cased borehole was similar as in the previous experiments. Additionally, the temperature increase in the TFW was applied faster than in the previous experiments in order to investigate the rock response to a heavier thermal load. It consists in a main borehole containing the heater and two instrumentation boreholes equipped with a temperature and a pore water pressure sensor each. The sensors installed in borehole TER1906 are at approximately 40 cm from the heater borehole wall in the bedding plane. The TER1907 sensors are in the direction perpendicular to the bedding planes at about 70 cm from the heater borehole. The boreholes were drilled from the GEX gallery in the direction of the major in situ stress (16 MPa). The heater has an effective heating length of 3.29 m and is located between 7 and 10 m from the GEX gallery. It was emplaced on a base to centre it in the borehole and to have a void between the heater and the rock mass for free convergence. Heating started on 10 January 2011 and lasted 56 days. One of the requirements of the experiment was to apply a relatively fast heating ramp in comparison with the previous tests in order to investigate the rock behaviour under stronger thermal load. The increase from 21 C to 90 C at the external heater wall was achieved in 14 hours. The initial pore water pressure is about 3.2 MPa, somewhat lower than the undisturbed pore water pressure at the level of the laboratory (4.5 MPa). This difference is attributed to the presence of the GEX gallery that was excavated in July-September 2005. In this work, we present the

Steady natural convection heat transfer experiments in a horizontal annulus for the United States Spent Fuel Shipping Cask Technology Program

International Nuclear Information System (INIS)

Boyd, R.D.

1981-04-01

This experimental study deals with the measurement of the heat transfer across a horizontal annulus which is formed by an inner hexagonal cylinder and an outer concentric circular cylinder. The geometry simulates, in two dimensions, a liquid metal fast breeder reactor radioactive fuel subassembly inside a shipping container. This geometry is also similar to a radioactive fuel pin inside a horizontal reactor subassembly. The objective of the experiments is to measure the local and mean heat transfer at the surface of the inner hexagonal cylinder

Heat cascading regenerative sorption heat pump

Science.gov (United States)

Jones, Jack A. (Inventor)

1995-01-01

A simple heat cascading regenerative sorption heat pump process with rejected or waste heat from a higher temperature chemisorption circuit (HTCC) powering a lower temperature physisorption circuit (LTPC) which provides a 30% total improvement over simple regenerative physisorption compression heat pumps when ammonia is both the chemisorbate and physisorbate, and a total improvement of 50% or more for LTPC having two pressure stages. The HTCC contains ammonia and a chemisorbent therefor contained in a plurality of canisters, a condenser-evaporator-radiator system, and a heater, operatively connected together. The LTPC contains ammonia and a physisorbent therefor contained in a plurality of compressors, a condenser-evaporator-radiator system, operatively connected together. A closed heat transfer circuit (CHTC) is provided which contains a flowing heat transfer liquid (FHTL) in thermal communication with each canister and each compressor for cascading heat from the HTCC to the LTPC. Heat is regenerated within the LTPC by transferring heat from one compressor to another. In one embodiment the regeneration is performed by another CHTC containing another FHTL in thermal communication with each compressor. In another embodiment the HTCC powers a lower temperature ammonia water absorption circuit (LTAWAC) which contains a generator-absorber system containing the absorbent, and a condenser-evaporator-radiator system, operatively connected together. The absorbent is water or an absorbent aqueous solution. A CHTC is provided which contains a FHTL in thermal communication with the generator for cascading heat from the HTCC to the LTAWAC. Heat is regenerated within the LTAWAC by transferring heat from the generator to the absorber. The chemical composition of the chemisorbent is different than the chemical composition of the physisorbent, and the absorbent. The chemical composition of the FHTL is different than the chemisorbent, the physisorbent, the absorbent, and ammonia.

The advanced containment experiments (ACE) Project

International Nuclear Information System (INIS)

Sehgal, B.R.; Ritzman, R.; Merilo, M.; Rahn, F.; Machiels, A.

1992-01-01

The overall structure and content of the ACE Project, which has been obtaining experimental data in four key areas of LWR severe accident technology are described. The key areas consist of filtration systems for vented containment concepts, radioiodine behavior in containment, the interaction of molten core material with structural concrete, and the use of water to terminate the core-concrete interaction process. Experiment procedures used in each phase of the work are summarized and the principal results and conclusions developed to date are discussed

Mixed convection heat transfer experiments using analogy concept

International Nuclear Information System (INIS)

Ko, Bong Jin; Chung, Bum Jin; Lee, Won Jea

2009-01-01

A Series of the turbulent mixed convective heat transfer experiments in a vertical cylinder was carried out. In order to achieve high Gr and/or Ra with small scale test rigs, the analogy concept was adopted. Using the concept, heat transfer systems were simulated by mass transfer systems, and large Grashof numbers could be achieved with reasonable facility heights. The tests were performed with buoyancy-aided flow and opposed flow for Reynolds numbers from 4,000 to 10,000 with a constant Grashof number, Gr H of 6.2 x 10 9 and Prandtl number of about 2,000. The test results reproduced the typical of the mixed convection heat transfer phenomena in a turbulent situation and agree well with the experimental study performed by Y. Palratan et al. The analogy experimental method simulated the mixed convection heat transfer phenomena successfully and seems to be a useful tool for heat transfer studies for VHTR as well as the systems with high buoyancy condition and high Prandtl number

Progress of High Heat Flux Component Manufacture and Heat Load Experiments in China

Energy Technology Data Exchange (ETDEWEB)

Liu, X.; Lian, Y.; Xu, Z.; Chen, J.; Chen, L.; Wang, Q.; Duan, X., E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, Chengu (China); Luo, G. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); Yan, Q. [University of Science and Technology Beijing, Beijing (China)

2012-09-15

Full text: High heat flux components for first wall and divertor are the key subassembly of the present fusion experiment apparatus and fusion reactors in the future. It is requested the metallurgical bonding among the plasma facing materials (PFMs), heat sink and support materials. As to PFMs, ITER grade vacuum hot pressed beryllium CN-G01 was developed in China and has been accepted as the reference material of ITER first wall. Additionally pure tungsten and tungsten alloys, as well as chemical vapor deposition (CVD) W coating are being developed for the aims of ITER divertor application and the demand of domestic fusion devices, and significant progress has been achieved. For plasma facing components (PFCs), high heat flux components used for divertor chamber are being studied according to the development program of the fusion experiment reactor of China. Two reference joining techniques of W/Cu mockups for ITER divertor chamber are being developed, one is mono-block structure by pure copper casting of tungsten surface following by hot iso-static press (HIP), and another is flat structure by brazing. The critical acceptance criteria of high heat flux components are their high heat load performance. A 60 kW Electron-beam Material testing Scenario (EMS-60) has been constructed at Southwestern Institute of Physics (SWIP),which adopts an electron beam welding gun with maximum energy of 150 keV and 150 x 150 mm{sup 2} scanning area by maximum frame rate of 30 kHz. Furthermore, an Engineering Mockup testing Scenario (EMS-400) facility with 400 kW electron-beam melting gun is under construction and will be available by the end of this year. After that, China will have the comprehensive capability of high heat load evaluation from PFMs and small-scale mockups to engineering full scale PFCs. A brazed W/CuCrZr mockup with 25 x 25 x 40 mm{sup 3} in dimension was tested at EMS-60. The heating and cooling time are 10 seconds and 15 seconds, respectively. The experiment

First Experience from the World Largest fully commercial Solar Heating Plant

DEFF Research Database (Denmark)

Heller, Alfred; Furbo, Simon

1997-01-01

The first experience from the largest solar heating plant in the world is given. The plant is situated in Marstal and is has a total area of 8000 square m.......The first experience from the largest solar heating plant in the world is given. The plant is situated in Marstal and is has a total area of 8000 square m....

Magnetic behaviors of cataclasites within Wenchuan earthquake fault zone in heating experiments

Science.gov (United States)

Zhang, L.; Li, H.; Sun, Z.; Chou, Y. M.; Cao, Y., Jr.; Huan, W.; Ye, X.; He, X.

2017-12-01

Previous rock magnetism of fault rocks were used to trace the frictional heating temperature, however, few studies are focus on different temperatures effect of rock magnetic properties. To investigate rock magnetic response to different temperature, we conducted heating experiments on cataclasites from the Wenchuan earthquake Fault Scientific Drilling borehole 2 (WFSD-2) cores. Samples of cataclasites were obtained using an electric drill with a 1 cm-diameter drill pipe from 580.65 m-depth. Experiments were performed by a Thermal-optical measurement system under argon atmosphere and elevated temperatures. Both microstructural observations and powder X-ray diffraction analyses show that feldspar and quartz start to melt at 1100 ° and 1300 ° respectively. Magnetic susceptibility values of samples after heating are higher than that before heating. Samples after heating at 700 and 1750 ° have the highest values of magnetic susceptibility. Rock magnetic measurements show that the main ferromagnetic minerals within samples heated below 1100 ° (400, 700, 900 and 1100 °) are magnetite, which is new-formed by transformation of paramagnetic minerals. The χferri results show that the quantity of magnetite is bigger at sample heated by 700° experiment than by 400, 900 and 1100° experiments. Based on the FORC diagrams, we consider that magnetite grains are getting finer from 400 to 900°, and growing coarser when heated from 900 to 1100 °. SEM-EDX results indicate that the pure iron are formed in higher temperature (1300, 1500 and 1750 °), which present as framboids with size values of samples when heated at 400, 700, 900 and 1100°, while the neoformed pure iron is responsible to the higher magnetic susceptibility values of samples when heated at 1300, 1500 and 1750°.

Weak Solution and Weakly Uniformly Bounded Solution of Impulsive Heat Equations Containing “Maximum” Temperature

Directory of Open Access Journals (Sweden)

Oyelami, Benjamin Oyediran

2013-09-01

Full Text Available In this paper, criteria for the existence of weak solutions and uniformly weak bounded solution of impulsive heat equation containing maximum temperature are investigated and results obtained. An example is given for heat flow system with impulsive temperature using maximum temperature simulator and criteria for the uniformly weak bounded of solutions of the system are obtained.

Blowdown heat transfer experiment, (1)

International Nuclear Information System (INIS)

Soda, Kunihisa; Yamamoto, Nobuo; Osaki, Hideki; Shiba, Masayoshi

1976-09-01

Blowdown heat transfer experiment has been carried out with a transparent test section to observe phenomena in coolant behavior during blowdown process. Experimental parameters are discharge position, initial system pressure, initial coolant temperature, power supply to heater rods and number of heater rods. At initial pressure 7-12 ata and initial power 6-50 kw per one heater rod, the flow condition in the test section is a major factor in determining time of DNB occurrence and physical process to DNB during blowdown. (auth.)

Hydride heat pump with heat regenerator

Science.gov (United States)

Jones, Jack A. (Inventor)

1991-01-01

A regenerative hydride heat pump process and system is provided which can regenerate a high percentage of the sensible heat of the system. A series of at least four canisters containing a lower temperature performing hydride and a series of at least four canisters containing a higher temperature performing hydride is provided. Each canister contains a heat conductive passageway through which a heat transfer fluid is circulated so that sensible heat is regenerated. The process and system are useful for air conditioning rooms, providing room heat in the winter or for hot water heating throughout the year, and, in general, for pumping heat from a lower temperature to a higher temperature.

Recent {sup 3}He radio frequency heating experiments on JET

Energy Technology Data Exchange (ETDEWEB)

Van Eester, D. [Association Euratom-Belgian State, LPP-ERM/KMS, TEC, Brussels (Belgium); Imbeaux, F.; Joffrin, E. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France)] [and others

2003-07-01

Various ITER relevant experiments using {sup 3}He in a majority D plasma were performed in the recent JET campaigns. Two types can be distinguished: dedicated studies of the various RF heating scenarios which rely on the presence of {sup 3}He, and physics studies using RF heating as a working tool to provide a tunable heat source. As the success of a number of these experiments depended on the capability to keep the {sup 3}He concentration fixed, real time control of the {sup 3}He concentration was developed and used. This paper presents a brief overview of the results obtained, zooms in on some of the more interesting recent findings and discusses some of the theoretical background. (authors)

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

International Nuclear Information System (INIS)

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

1994-08-01

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

Impact of bulk atmospheric motion on local and global containment heat transfer

International Nuclear Information System (INIS)

Green, J.A.; Almenas, K.

1995-01-01

Local and global correlations for condensing energy transfer in the presence of noncondensable gases in a containment facility have been evaluated. The database employed stems from the E11.2 and E11.4 tests conducted at the German HDR facility. The HDR containment is a 11060-ml, 60-m-high decommissioned light water reactor. The tests simulated long-term (up to 56 h) accident conditions. Numerous instrumented structural blocks (concrete and lead) were located throughout the containment to provide detailed local heat transfer measurements. These data represent what is probably the most extensive database of integral energy transfer measurements available. It is well established that the major resistance to condensation heat transfer in the presence of noncondensable gases is a gaseous boundary layer that builds up in front of the condensing surface. Correlations that seek to model heat transfer for these conditions should depend on parameters that most strongly determine the buildup and thickness of this boundary layer. Two of the most important parameters are the vapor/noncondensable concentration ratio and the local atmospheric motion. Secondary parameters include the atmosphere-to-surface temperature difference, the pressure, and condensing surface properties. The HDR tests are unique in terms of the quantity and variety of instrumentation employed. However, one of the most important parameters, the local bulk atmospheric velocity, is inherently difficult to measure, and only fragmentary measurements are available even in the HDR data-base. A detailed analysis of these data is presented by Green. This study uses statistical methods to evaluate local and global empirical correlations that do not include the atmospheric velocity. The magnitude of the differences between the correlations emphasizes the importance of the local atmospheric velocity and serves to illustrate the accuracy limits of correlations that neglect this essential parameter

Aerodynamics, heat and mass transfer in steam-aerosol turbulent flows in containment

Energy Technology Data Exchange (ETDEWEB)

Nigmatulin, B.I.; Pershukov, V.A.; Ris, V.V. [Research & Engineering Centre of Nuclear Plants Safety, Moscow (Russian Federation)] [and others

1995-09-01

In this report an analysis of aerodynamic and heat transfer processes at the blowdown of gas-dispersed mixture into the containment volume is presented. A few models for description of the volume averaged and local characteristics are analyzed. The mathematical model for description of the local characteristics of the turbulent gas-dispersed flows was developed. The calculation of aerodynamic, heat and mass transfer characteristics was based on the Navier-Stokes, energy and gas mass fractions conservation equations. For calculation of dynamics and deposition of the aerosols the original diffusion-inertia model is developed. The pulsating characteristics of the gaseous phase were calculated on the base (k-{xi}) model of turbulence with modification to account thermogravitational force action and influence of particle mass loading. The appropriate boundary conditions using the {open_quotes}near-wall function{close_quotes} approach was obtained. Testing of the mathematical models and boundary conditions has shown a good agreement between computation and data of comparison. The described mathematical models were applied to two- and three dimensional calculations of the turbulent flow in containment at the various stages of the accident.

Household preferences of hybrid home heating systems – A choice experiment application

International Nuclear Information System (INIS)

Ruokamo, Enni

2016-01-01

The residential heating sector presents considerable energy savings potential, as numerous heating solutions for reducing electricity consumption and utilizing renewable energy sources are available in the market. The aim of this paper is to examine determinants of household heating system choices and to use this information for policy planning purposes. This paper investigates residential homeowner attitudes regarding innovative hybrid home heating systems (HHHS) with choice experiment. Heating system scenarios are designed to represent the most relevant primary and supplementary heating alternatives currently available in Finland. The choice sets include six main heating alternatives (district heat, solid wood, wood pellet, electric storage heating, ground heat pump and exhaust air heat pump) that are described by five attributes (supplementary heating systems, investment costs, operating costs, comfort of use and environmental friendliness). The results imply that HHHSs generally appear to be accepted among households; however, several factors affect perceptions of these technologies. The results reveal differing household attitudes toward the main heating alternatives and show that such views are affected by socio-demographic characteristics (age, living environment, education, etc.). The results suggest that households view supplementary heating systems (especially solar-based) favorably. The other attributes studied also play a significant role in decision making. - Highlights: •Study of hybrid heating where supplementary and main heating systems are combined. •Choice experiment is applied to study the determinants of hybrid heating adoption. •Hybrid heating appears to be generally accepted among households. •Households exhibit differing attitudes toward hybrid heating. •Policy makers should not underestimate the potential of hybrid heating.

CONTAIN assessment of the NUPEC mixing experiments

International Nuclear Information System (INIS)

Stamps, D.W.

1995-08-01

The ability of the CONTAIN code to predict the thermal hydraulics of five experiments performed in the NUPEC 1/4-scale model containment was assessed. These experiments simulated severe accident conditions in a nuclear power plant in which helium (as a nonflammable substitute for hydrogen) and steam were coinjected at different locations in the facility with and without the concurrent injection of water sprays in the dome. Helium concentrations, gas temperatures and pressures, and wall temperatures were predicted and compared with the data. The use of different flow solvers, nodalization schemes, and analysis methods for the treatment of water sprays was emphasized. As a result, a general procedure was suggested for lumped-parameter code analyses of problems in which the thermal hydraulics are dominated by water sprays

Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

International Nuclear Information System (INIS)

Powell, J.; Reich, M.; Barletta, R.

1996-01-01

A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

Progress on Electron Cyclotron Heating Experiments in LHD

International Nuclear Information System (INIS)

Shimozuma, T.; Kubo, S.; Yoshimura, Y.; Igami, H.; Nagasaki, K.; Notake, T.; Inagaki, S.; Ito, S.; Kobayashi, S.; Mizuno, Y.; Takita, Y.; Ohkubo, K.; Saito, K.; Seki, T.; Kumazawa, R.; Watari, T.; Mutoh, T.

2005-01-01

Electron cyclotron resonance heating (ECH) is a powerful heating method because of its well-controlled local heating and high deposition power density. Together with the development of high power long pulse gyrotrons, ECH becomes one of the major heating scenarios to control electron temperature and current profiles for the improved plasma confinement and suppression of some magneto-hydro-dainamic (MHD) instabilities in both tokamaks and stellarators [1]. In the Large Helical Device (LHD), ECH has been worked as a method of plasma initiation and electron heating. The ECH system has been improved with respect to each experimental campaign. In the recent campaign, nine gyrotrons were operated reliably and steadily. As a diagnostic objective, a modulated ECH (MECH) was injected together with main ECH power. A Fourier analysis of the induced heat wave gave useful information of not only the heat transport in the plasmas but also precise power deposition layer [2]. Several kinds of ECH experiment were performed by using this flexible ECH system. In LHD, electron ITB formation have been observed by using strongly focused ECH in the plasma core [3].Two different kinds of improved confinement were realized depending on the direction of tangentially injected NBI. NBI beam driven currents modify the profiles of the rotational transform 2 ro and the existence low order rational surfaces, 2 = 0.5 in special, affects the difference of appearance of the improved confinement states. The MECH method was used to investigate the internal structure of the thermal diffusion in such plasmas [4]. Another important role of the MECH is the precise determination of the ECH power deposition. Shift of the deposition location by changing an injection polarization in the electron Bernstein wave (EBW) heating was clearly demonstrated by the MECH method. Electron cyclotron current drive (ECCD) experiments were proceeded by using a flexible antenna system, which had wide scanning range in both

Seasonal distributions of diabatic heating during the First GARP Global Experiment

OpenAIRE

Ying Wei, Ming; Johnson, Donald R.; Townsend, Ronald D.

2011-01-01

The seasonal and annual global distributions of diabatic heating during the First GARP Global Experiment (FGGE) are estimated using the isentropic mass continuity equation. The data used are from the FGGE Level IIIa analyses generated by the United States National Meteorological Center. Spatially and temporally coherent diabatic heating distributions are obtained from the isentropic planetary scale mass circulation that is forced by large-scale heat sources and sinks. The diabatic heating in...

Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

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

1991-01-01

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

Radiant heat testing of the H1224A shipping/storage container

Energy Technology Data Exchange (ETDEWEB)

Harding, D.C.; Bobbe, J.G.; Stenberg, D.R.; Arviso, M.

1994-05-01

H1224A weapons containers have been used for years by the Departments of Energy and Defense to transport and store W78 warhead midsections. Although designed to protect the midsections only from low-energy impacts, a recent transportation risk assessment effort has identified a need to evaluate the container`s ability to protect weapons in more severe accident environments. Four radiant heat tests were performed: two each on an H1224A container (with a Mk12a Mod 6c mass mock-up midsection inside) and two on a low-cost simulated H1224A container (with a hollow Mk12 aeroshell midsections inside). For each unit tested, temperatures were recorded at numerous points throughout the container and midsection during a 4-hour 121{degrees}C (250{degrees}F) and 30-minute 1010{degrees}C (1850{degrees}F) radiant environment. Measured peak temperatures experienced by the inner walls of the midsections as a result of exposure to the high-temperature radiant environment ranged from 650{degrees} C to 980{degrees} C (1200{degrees} F to 1800{degrees}F) for the H1224A container and 770 {degrees} to 990 {degrees}C (1420{degrees} F to 1810{degrees}F) for the simulated container. The majority of both containers were completely destroyed during the high-temperature test. Temperature profiles will be used to benchmark analytical models and predict warhead midsection temperatures over a wide range of the thermal accident conditions.

Imitation experiment for water-treatment by heat of solar collector and hot pump

International Nuclear Information System (INIS)

Liao Yuanzong; Liu Shuqing; Pang Heding; Zhao Zhongxin; Zhang Biguang; Wang Xiping; Huo Guangqing

1997-01-01

The author presents an imitation experiment in which solar collector and hot pump are jointed for supplying heat to evaporate cleaned water and diffuse it into air. The effects of the temperature and the quantity of supplying air, and circumstance conditions on evaporation quantity are studied. The ratio of evaporating quantity to consuming energy, the efficiency of evaporation, average efficiency of solar collector and supplying heat coefficient of heat pump are measured. The experiment shows that this supplying heat model is practicable, economic and efficient for treating cleaned water

Magnetic power conversion with machines containing full or porous wheel heat exchangers

Science.gov (United States)

Egolf, Peter W.; Kitanovski, Andrej; Diebold, Marc; Gonin, Cyrill; Vuarnoz, Didier

2009-04-01

A first part of the article contains a thermodynamic theory describing the temperature distribution in a Curie wheel. The occurring nonlinear ordinary differential equation has an analytical solution. If a Curie wheel is stabilized by levitation, it is named Palmy wheel. These wheels show a full structure, and because of this reason, their uptake of heat from a flame (Curie wheel) or by (solar) light absorption (Palmy wheel) only on the periphery of a cylinder is very limited. To improve the method, a modification of the principle by introducing a convective heat transport into a porous wheel is discussed. By this the power conversion rate from a heat flux to mechanical and electric power is very much increased. The second part of the article presents results of a theoretical/numerical study on the efficiencies of magnetic power conversion plants operating with porous wheels. Furthermore, these efficiencies—which are promising—are compared with those of existing power conversion plants, as e.g. geothermal binary cycle power plants.

Magnetic power conversion with machines containing full or porous wheel heat exchangers

International Nuclear Information System (INIS)

Egolf, Peter W.; Kitanovski, Andrej; Diebold, Marc; Gonin, Cyrill; Vuarnoz, Didier

2009-01-01

A first part of the article contains a thermodynamic theory describing the temperature distribution in a Curie wheel. The occurring nonlinear ordinary differential equation has an analytical solution. If a Curie wheel is stabilized by levitation, it is named Palmy wheel. These wheels show a full structure, and because of this reason, their uptake of heat from a flame (Curie wheel) or by (solar) light absorption (Palmy wheel) only on the periphery of a cylinder is very limited. To improve the method, a modification of the principle by introducing a convective heat transport into a porous wheel is discussed. By this the power conversion rate from a heat flux to mechanical and electric power is very much increased. The second part of the article presents results of a theoretical/numerical study on the efficiencies of magnetic power conversion plants operating with porous wheels. Furthermore, these efficiencies-which are promising-are compared with those of existing power conversion plants, as e.g. geothermal binary cycle power plants

Laser heated solenoid proof-of-concept experiment (PCX) facility

International Nuclear Information System (INIS)

DeHart, T.E.; Zumdieck, J.F.; Hoffman, A.L.; Lowenthal, D.D.; Crawford, E.A.; Parry, B.

1977-01-01

The total facility, including laser, magnet, focusing optics, instrumentation and control, its design problems, and its current performance are discussed. Preliminary results from plasma heating experiments are discussed

High frequency ion Bernstein wave heating experiment on JIPP T-IIU tokamak

International Nuclear Information System (INIS)

Seki, T.; Kumazawa, R.; Watari, T.

1992-08-01

An experiment in a new regime of ion Bernstein wave (IBW) heating has been carried out using 130 MHz high power transmitters in the JIPP T-IIU tokamak. The heating regime utilized the IBW branch between the 3rd and 4th harmonics of the hydrogen ion cyclotron frequencies. This harmonic number is the highest among those used in the IBW experiments ever conducted. The net radio-frequency (RF) power injected into the plasma is around 400 kW, limited by the transmitter output power. Core heating of ions and electrons was confirmed in the experiment and density profile peaking was found to feature the IBW heating (IBWH). The peaking of the density profile was also found when IBW was applied to the neutral beam injection heated discharges. An analysis by use of a transport code with these experimental data indicates that the particle confinement should be improved in the plasma core region on the application of IBWH. It is also found that the ion energy distribution function observed during IBWH has less high energy tail than those in conventional ion cyclotron range of frequency heating regimes. The observed IBWH-produced ion energy distribution function is in a reasonable agreement with the calculation based on the quasi-linear RF diffusion / Fokker-Planck model. (author)

Effect of seasonal variation on some physical properties and heat stability of milk subjected to ultra-high temperature and in-container sterilisation.

Science.gov (United States)

Chen, Biye; Grandison, Alistair S; Lewis, Michael J

2015-08-15

Heat stability was evaluated in bulk raw milk, collected throughout the year and subjected to ultra-high temperature (UHT) or in-container sterilisation, with and without added calcium chloride (2mM), disodium hydrogen phosphate (DSHP, 10mM) and trisodium citrate (TSC, 10mM). More sediment was observed following in-container sterilisation (0.24%) compared with UHT (0.19%). Adding CaCl2 made the milk more unstable to UHT than to in-container sterilisation, while adding DSHP and TSC made the milk more unstable during in-container sterilisation than to UHT processing, although TSC addition increased the sediment formed by UHT processing. Better heat stability was observed in autumn and winter than in spring and summer following UHT. However, following in-container sterilisation, samples with added stabilising salts showed significantly improved heat stability in autumn, whereas with added CaCl2, the best heat stability was observed in spring. No correlation was found between urea and heat stability. Copyright © 2015 Elsevier Ltd. All rights reserved.

On heat transfer characteristics of real and simulant melt pool experiments

Energy Technology Data Exchange (ETDEWEB)

Dinh, T.N.; Nourgaliev R.R.; Sehgal, B.R. [Royal Institute of Technology, Stockholm (Sweden)

1995-09-01

The paper presents results of analytical studies of natural convection heat transfer in scaled and/or simulant melt pool experiments related to the PWR in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing physical processes under prototypical situations is examined. Experiments and experimental approaches are analysed by focusing on their ability to represent prototypical situations. Calculations are carried out in order to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. Rayleigh numbers in the present analysis are limited to 10{sup 12}, where uncertainties in turbulence modeling are not overriding other uncertainties. The effects of fluid Prandtl number on heat transfer to the lowermost part of cooled pool walls are examined for square and semicircular cavities. Calculations are performed also to explore limitations of using side-wall heating and direct electrical heating in reproducing the physical picture of interest. Needs for further experimental and analytical efforts are discussed as well.

LMFBR intermediate-heat-exchanger experience

International Nuclear Information System (INIS)

Cho, S.M.; Beaver, T.R.

1983-01-01

This paper presents developmental and operating experience of large Intermediate Heat Exchangers (IHX's) in US from the Fast Flux Test Facility (FFTF) to the Clinch River Breeder Reactor Plant (CRBRP) to the Large Development Plant (LDP). Design commonalities and deviations among these IHX's are synopsized. Various developmental tests that were conducted in the areas of hydraulic, structural and mechanical design are also presented. The FFTF is currently operating. Performance data of the FFTF IHXs are reviewed, and comparisons between actual and predicted performances are made. The results are used to assess the adequacy of IHX designs

The use of ferrofluids for heat removal: Advantage or disadvantage?

Energy Technology Data Exchange (ETDEWEB)

Krauzina, Marina T., E-mail: krauzina@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Bozhko, Aleksandra A., E-mail: bozhko@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Krauzin, Pavel V., E-mail: krauzin@psu.ru [Faculty of Physics, Perm State University, 15 Bukirev Street, Perm 614990 (Russian Federation); Suslov, Sergey A., E-mail: ssuslov@swin.edu.au [Department of Mathematics H38, Swinburne University of Technology, Hawthorn, Victoria 3122 (Australia)

2017-06-01

It is shown experimentally that, depending on the relative orientation of the gravity and the thermal gradient and on the pre-history of experiment, the application of a uniform external vertical magnetic field to a spherical cavity filled with magnetic ferrofluid can either enhance or suppress a convective heat transfer. - Highlights: • Conduction heat transfer in magnetic fluid heated from above is stronger than that in a fluid not containing nanoparticles. • The application of a uniform vertical magnetic field enhances heat transfer when magnetic fluid is heated from above. • Heat transfer in a magnetic fluid heated from below is weaker than that in a fluid not containing nanoparticles.

Simulation and experiment on the thermal performance of U-vertical ground coupled heat exchanger

Energy Technology Data Exchange (ETDEWEB)

Li, Xinguo; Chen, Zhihao; Zhao, Jun [Department of Thermal Engineering, School of Mechanical Engineering, Tianjin University, Tianjin 300072 (China)

2006-10-15

This paper presented both the numerical simulations and experiments on the thermal performance of U-vertical ground coupled heat exchanger (UGCHE). The variation of the ground temperature and heat balance of the system were analyzed and compared in different operation modes in the numerical simulation. Experiments on the operation performance of the ground-coupled heat pump (GCHP) with the UGCHE were carried out. It shows that the ground source can be used as the heat source/sink for GCHP systems to have higher efficiency in saving energy. To preserve the ground resource for the sustainable utilization as heat source/sink, the heat emitted to ground and heat extracted from ground should be balanced. (author)

Experiments and simulations on heat exchangers in thermoelectric generator for automotive application

International Nuclear Information System (INIS)

Liu, X.; Deng, Y.D.; Zhang, K.; Xu, M.; Xu, Y.; Su, C.Q.

2014-01-01

In this work, an energy-harvesting system which extracts heat from an automotive exhaust pipe and turns the heat into electricity by using thermoelectric power generators (TEGs) was built. Experiments show that the temperature difference in automotive system is not constant, especially the heat exchanger, which cannot provide the thermoelectric modules (TMs) large amount of heat. The thermal performance of different heat exchangers in exhaust-based TEGs is studied in this work, and the thermal characteristics of heat exchangers with different internal structures and thickness are discussed, to obtain higher interface temperature and thermal uniformity. Following computational fluid dynamics simulations, infrared experiments and output power testing system are carried out on a high-performance production engine with a dynamometer. Results show that a plate-shaped heat exchanger with chaos-shaped internal structure and thickness of 5 mm achieves a relatively ideal thermal performance, which is practically useful to enhance the thermal performance of the TEG, and larger total output power can be thus obtained. - Graphical abstract: The thermal and electrical characteristics of different heat exchangers of automotive exhaust-based thermoelectric generator are discussed, to obtain higher interface temperature and thermal uniformity. - Highlights: • Different internal structures and thickness of heat exchangers were proposed. • Power output testing system of the two heat exchangers was characterized. • Chaos-shaped heat exchanger (5 mm thickness) shows better performance

The ISHTE [In-Situ Heat Transfer Experiment] lander: Final report

International Nuclear Information System (INIS)

Olson, L.O.; Harrison, J.G.

1986-12-01

This report describes the design and development of a sea floor lander constructed to support the In-Situ Heat Transfer Experiment (ISHTE). The work entailed fabricating and testing a steel space frame that would support and accurately position delicate instruments which would monitor a heat source driven into the sediments of the deep ocean. This lander is capable of being (1) transported from Seattle to Hawaii and back several times; (2) deployed from a ship at sea; (3) operated on the sea floor to field delicate experimental equipment; and (4) recovered for retrofit to support a one-year experiment on the sea floor

Heat Transfer Modes and their Coefficients for a Passive Containment Cooling System of PWR using a Multi-Pod Heat Pipe

Energy Technology Data Exchange (ETDEWEB)

Nam, Gyeongho; Park, Junseok; Kim, Sangnyung [Kyung Hee Univ., Yongin (Korea, Republic of)

2013-05-15

If a reactor core is damaged due to a disaster such as happened at TEPCO's Fukushima nuclear power plant, the inevitable rise of super-heated steam that could potentially convert to hydrogen resulting from unimpeded temperature and pressure rises will threaten the integrity of the containment structure. To prevent this, safety and regulatory standards typically specify that the gas vent and external cooling systems be designed to maintain containment up to the level C limit for 24 hours and integrity for 48 hours after any damage to the core. Furthermore, it is recommended that the installation of the exhaust penetration unit have a minimum diameter of 3ft. However, installation of such cooling measures or penetration units is burdensome in terms of operational and maintenance costs not to mention the need to ensure a fleet of fire trucks to be on standby as well as the need to ensure a plentiful supply of water for cooling and a filtration system to clean the water. Therefore, the development of a reliable passive cooling system will be economically advantageous because the extra cost burdens of the external system can be omitted. The Passive Containment Cooling System (PCCS) using a multi-pod heat pipe proposed in this study satisfies these conditions.

Critical heat flux experiments in tight lattice core

Energy Technology Data Exchange (ETDEWEB)

Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-12-01

Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

Critical heat flux experiments in tight lattice core

International Nuclear Information System (INIS)

Kureta, Masatoshi

2002-01-01

Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

Experiment on heat transfer in simulated molten core/concrete interaction

International Nuclear Information System (INIS)

Katsumura, Yukihiro; Hashizume, Hidetoshi; Toda, Saburo; Kawaguchi, Takahiro.

1993-01-01

In order to investigate heat transfer between molten core and concrete in LWR severe accidents, experiments were performed using water as the molten core, paraffin as the concrete, and air as gases from the decomposition of concrete. It was found that the heat transfer on the interface between paraffin and water were promoted strongly by the air gas. (author)

Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal

International Nuclear Information System (INIS)

Stehle, H.; Damm, G.; Jansing, W.

1994-01-01

Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA. (orig.)

Restoration of the Apollo Heat Flow Experiments Metadata

Science.gov (United States)

Nagihara, S.; Stephens, M. K.; Taylor, P. T.; Williams, D. R.; Hills, H. K.; Nakamura, Y.

2015-01-01

Geothermal heat flow probes were deployed on the Apollo 15 and 17 missions as part of the Apollo Lunar Surface Experiments Package (ALSEP). At each landing site, the astronauts drilled 2 holes, 10-m apart, and installed a probe in each. The holes were 1- and 1.5-m deep at the Apollo 15 site and 2.5-m deep at the Apollo 17 sites. The probes monitored surface temperature and subsurface temperatures at different depths. At the Apollo 15 site, the monitoring continued from July 1971 to January 1977. At the Apollo 17 site, it did from December 1972 to September 1977. Based on the observations made through December 1974, Marcus Langseth, the principal investigator of the heat flow experiments (HFE), determined the thermal conductivity of the lunar regolith by mathematically modeling how the seasonal temperature fluctuation propagated down through the regolith. He also determined the temperature unaffected by diurnal and seasonal thermal waves of the regolith at different depths, which yielded the geothermal gradient. By multiplying the thermal gradient and the thermal conductivity, Langseth obtained the endogenic heat flow of the Moon as 21 mW/m(exp 2) at Site 15 and 16 mW/m(exp 2) at Site 17.

State of the art on the heat transfer experiments under supercritical pressure condition

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Song, Chul Hwa

2003-07-01

The SCWR(Super-Critical Water cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project which aims at the development of new reactors with enhanced economy and safety. The SCWR is considered to be a feasible concept of new nuclear power plant if the existing technologies developed in fossil fuel fired plant and LWR technologies together with additional research on several disciplines such as materials, water chemistry and safety. As KAERI takes part in the GIF(Generation IV Forum) for the Gen-IV project, domestic concerns about the SCWR have been recently increased. In order to establish a foundation for the development of SCWR, efforts should be concentrated on the conceptual design of systems and the associated key experiments as well. Heat transfer experiments, among others, under supercritical condition are required for the proper prediction of thermal hydraulic phenomena, which are essential for the thermal hydraulic designs of reactor core. Nevertheless, the experiments have not been performed in Korea yet. This report deals with fundamental surveys on the heat transfer experiments under supercritical conditions, which are required for the understanding of heat transfer characteristics for the thermal hydraulic designs of supercritical reactor core. Investigations on the physical properties of water and CO 2 showed that the physical properties such as density, specific heat, viscosity and thermal conductivity are significantly changed near the pseudo-critical points. The state of the art on the heat transfer characteristics in relation with heat transfer deterioration and heat transfer coefficient is briefly described. In addition, previous experiments with supercritical water as well as supercritical CO 2 and Freon used for an alternating fluid are presented

State of the art on the heat transfer experiments under supercritical pressure condition

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwan Yeol; Song, Chul Hwa

2003-07-01

The SCWR(Super-Critical Water cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project which aims at the development of new reactors with enhanced economy and safety. The SCWR is considered to be a feasible concept of new nuclear power plant if the existing technologies developed in fossil fuel fired plant and LWR technologies together with additional research on several disciplines such as materials, water chemistry and safety. As KAERI takes part in the GIF(Generation IV Forum) for the Gen-IV project, domestic concerns about the SCWR have been recently increased. In order to establish a foundation for the development of SCWR, efforts should be concentrated on the conceptual design of systems and the associated key experiments as well. Heat transfer experiments, among others, under supercritical condition are required for the proper prediction of thermal hydraulic phenomena, which are essential for the thermal hydraulic designs of reactor core. Nevertheless, the experiments have not been performed in Korea yet. This report deals with fundamental surveys on the heat transfer experiments under supercritical conditions, which are required for the understanding of heat transfer characteristics for the thermal hydraulic designs of supercritical reactor core. Investigations on the physical properties of water and CO{sub 2} showed that the physical properties such as density, specific heat, viscosity and thermal conductivity are significantly changed near the pseudo-critical points. The state of the art on the heat transfer characteristics in relation with heat transfer deterioration and heat transfer coefficient is briefly described. In addition, previous experiments with supercritical water as well as supercritical CO{sub 2} and Freon used for an alternating fluid are presented.

Study on solar chemical heat pump system. Basic experiment on dehydrogenation of 2-propanol using heteropoly-acid photo catalyst; Solar chemical heat pump no kenkyu. Heteropoly sankei hikari shokubai wo mochiita 2-propanol no dassuiso hanno

Energy Technology Data Exchange (ETDEWEB)

Doi, T [Electrochemical Laboratory, Tsukuba (Japan); T-Raissi, A; Muradov, N [Florida Solar Energy Center, FL (United States)

1996-10-27

With the purpose of converting solar heat energy to an industrial heat energy, an examination was carried out empirically on the case of using a heteropoly-acid photo catalyst for the decomposition reaction process of 2-propanol. The experiment was performed in Florida Solar Energy Center, in the U.S.A.. The device for the experiment was constituted of a reaction part, distribution manifold for feeding from the lower part of the reaction part a 2-propanol solution for which a photo catalyst was suspended, storage tank served also as a gas-liquid separating container, and circulating pump. Silica-tangstic acid was used as the photo catalyst. In an outdoor experiment using solar radiation, the quantity of inclined global solar radiation was 530-950W/m{sup 2} in clear days and 100-600W/m{sup 2} in cloudy days, with temperatures between 17 and 26{degree}C throughout the experiment period. In addition, an indoor experiment was also conducted using an artificial light source (UV light). As a result of the experiment, the energy conversion efficiency was at most about 1% of incident UV light, a low figure compared to a heat utilization ratio of approximately 15% with a thermal catalyst. 6 refs., 8 figs.

Investigation on chemical heat pump using calcium-chloride; Enka calcium no suiwa dassui hanno wo mochiita solar chemical heat pump ni kansuru kenkyu

Energy Technology Data Exchange (ETDEWEB)

Fujii, I; Arai, T; Saito, Y [Meiji University, Tokyo (Japan)

1997-11-25

With an objective of developing a room heating system utilizing a solar chemical heat pump, an experimental system was fabricated to evaluate its performance. Steam was employed as a working gas, and for a reaction material, calcium-chloride was used, which has a reaction temperature zone permitting safe use and fitting the purpose among other hydrate systems and has high standard enthalpy in hydration. Water was used as a solar heat transferring medium. The system operates under the following principle: a container I is filled with hydrated salt and a container II with water, the two containers being linked with a pipe interposed with a valve; heat is inputted and outputted by performing charging and discharging alternately; and the role of a heat pump is played by deriving from environment the heat of water evaporation in the container II during discharging. The COP must take into account the electric power consumption of the water circulation pump to transfer solar heat. A COP of 0.256 was derived as a result of the experiment. 3 refs., 5 figs.

Fuel coolant interaction experiment by direct electrical heating method

International Nuclear Information System (INIS)

Takeda, Tsuneo; Hirano, Kenmei

1979-01-01

In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

Chemistry experiences from a containment fire at Ringhals unit 2

International Nuclear Information System (INIS)

Arvidsson, Bengt; Svanberg, Pernilla; Bengtsson, Bernt

2012-09-01

containment, together with 1000 smear test (cotton pads) for chloride analysis in the chemistry laboratory to evaluate contamination levels and verify the cleaning procedures and results. The main chemistry issues and concerns have been related to surface and water contamination of chloride, bromide, carbon, lead, copper and zinc from corrosion point of view. Lack of specification and guidelines for several of this parameters forced Ringhals to establish some internal guidelines and technical basis for clean up and restart of the plant. The solubility of soot particles was found to be very low and more adhesive to surfaces at high temperature, this caused some concerns and actions to clean up reactor coolant from soot particles before fuel reload and heating. An extensive review of stainless steel Outer Diameter Stress Corrosion Cracking (ODSCC) was performed independently from the fire incident during the outage, indicating a high number of crack indications of 1-3 mm depth, all within acceptance criteria for material thickness and operation. The indications are more likely to be addressed to almost 40 years of operation in marine atmosphere then the fire itself, even if the chloride contamination from fire may have supported some propagation. All found cracks were grinded according to authority requirements and no pipes needed to be replaced. The heating and start-up of Ringhals 2 could be done successfully without any water chemistry deviations due to the fire and the following cycle have been normal. The cleanness of R2 containment surfaces are now highly improved compared to earlier outages or other sea-cooled power plants. However, an extended program has been introduced to follow external surface chloride contamination built up in containment more frequently, together with inspections of ODSCC. The workload from the containment fire has been extreme and the chemistry and corrosion experiences several. This paper gives a summary of the results, challenges, solutions and

Design of an experiment to measure the fire exposure of radioactive materials packages aboard container cargo ships

International Nuclear Information System (INIS)

Koski, J.A.

1997-11-01

The test described in this paper is intended to measure the typical accident environment for a radioactive materials package in a fire aboard a container cargo ship. A stack of nine used standard cargo containers will be variously loaded with empty packages, simulated packages and combustible cargo and placed over a large hydrocarbon pool fire of one hour duration. Both internal and external fire container fire environments typical of on-deck stowage will be measured as well as the potential for container to container fire spread. With the use of the inverse heat conduction calculations, the local heat transfer to the simulated packages can be estimated from thermocouple data. Data recorded will also provide information on fire durations in each container, fire intensity and container to container fire spread characteristics

Heat pipes theory, design and applications

CERN Document Server

Reay, David; Kew, Peter

2013-01-01

Heat Pipes, 6th Edition, takes a highly practical approach to the design and selection of heat pipes, making it an essential guide for practicing engineers and an ideal text for postgraduate students. This new edition has been revised to include new information on the underlying theory of heat pipes and heat transfer, and features fully updated applications, new data sections, and updated chapters on design and electronics cooling. The book is a useful reference for those with experience and an accessible introduction for those approaching the topic for the first time. Contains all informat

Full-scale and time-scale heating experiments at Stripa: preliminary results. Technical project report No. 11

International Nuclear Information System (INIS)

Cook, N.G.W.; Hood, M.

1978-12-01

Two full-scale heating experiments and a time-scale heating experiment have recently been started in granite 340 meters below surface. The purpose of the full-scale heating experiments is to assess the near-field effects of thermal loading for the design of an underground repository of nuclear wastes. That of the time-scale heating experiments is to obtain field data of the interaction between heaters and its effect on the rock mass during a period of about two years, which corresponds to about twenty years of full-scale operation. Geological features of the rock around each experiment have been mapped carefully, and temperatures, stresses and displacements induced in the rock by heating have been calculated in advance of the experiments. Some 800 different measurements are recorded at frequent intervals by a computer system situated underground. These data can be compared at any time with predictions made earlier on video display units underground

Review of direct electrical heating experiments on irradiated mixed-oxide fuel

International Nuclear Information System (INIS)

Fenske, G.R.; Bandyopadhyay, G.

1982-01-01

Results of approximately 50 out-of-reactor experiments that simulated various stages of a loss-of-flow event with irradiated fuel are presented. The tests, which utilized the direct electrical heating technique to simulate nuclear heating, were performed either on fuel segments with their original cladding intact or on fuel segments that were extruded into quartz tubes. The test results demonstrated that the macro- and microscopic fuel behavior was dependent on a number of variables including fuel heating rate, thermal history prior to a transient, the number of heating cycles, type of cladding (quartz vs stainless steel), and fuel burnup

Space chamber experiments of ohmic heating by high power microwave from the solar power satellite

Energy Technology Data Exchange (ETDEWEB)

Kaya, N.; Matsumoto, H.

1981-12-01

It is quantitatively predicted that a high power microwave from the Solar Power Satellite (SPS) nonlinearly interacts with the ionospheric plasma. The possible nonlinear interactions are ohmic heating, self-focusing and parametric instabilities. A rocket experiment called MINIX (Microwave-Ionosphere Nonlinear Interaction Experiment) has been attempted to examine these effects, but is note reported here. In parallel to the rocket experiment, a laboratory experiment in a space plasma simulation chamber has been carried out in order to examine ohmic heating in detail and to develop a system of the rocket experiment. Interesting results were observed and these results were utilized to revise the system of the rocket experiments. A significant microwave heating of plasma up to 150% temperature increase was observed with little electron density decrease. It was shown that the temperature increase is not due to the RF breakdown but to the ohmic heating in the simulated ionospheric plasma. These microwave effects have to be taken into account in the SPS Project in the future.

Operational experience with SLAC's beam containment electronics

International Nuclear Information System (INIS)

Constant, T.N.; Crook, K.; Heggie, D.

1977-03-01

Considerable operating experience was accumulated at SLAC with an extensive electronic system for the containment of high power accelerated beams. Average beam power at SLAC can approach 900 kilowatts with the potential for burning through beam stoppers, protection collimators, and other power absorbers within a few seconds. Fast, reliable, and redundant electronic monitoring circuits have been employed to provide some of the safeguards necessary for minimizing the risk to personnel. The electronic systems are described, and the design philosophy and operating experience are discussed

Modeling of high power ICRF heating experiments on TFTR

International Nuclear Information System (INIS)

Phillips, C.K.; Wilson, J.R.; Bell, M.; Fredrickson, E.; Hosea, J.C.; Majeski, R.; Ramsey, A.; Rogers, J.H.; Schilling, G.; Skinner, C.; Stevens, J.E.; Taylor, G.; Wong, K.L.; Murakami, M.

1993-01-01

Over the past two years, ICRF heating experiments have been performed on TFTR in the hydrogen minority heating regime with power levels reaching 11.2 MW in helium-4 majority plasmas and 8.4 MW in deuterium majority plasmas. For these power levels, the minority hydrogen ions, which comprise typically less than 10% of the total electron density, evolve into la very energetic, anisotropic non-Maxwellian distribution. Indeed, the excess perpendicular stored energy in these plasmas associated with the energetic minority tail ions is often as high as 25% of the total stored energy, as inferred from magnetic measurements. Enhanced losses of 0.5 MeV protons consistent with the presence of an energetic hydrogen component have also been observed. In ICRF heating experiments on JET at comparable and higher power levels and with similar parameters, it has been suggested that finite banana width effects have a noticeable effect on the ICRF power deposition. In particular, models indicate that finite orbit width effects lead to a reduction in the total stored energy and of the tail energy in the center of the plasma, relative to that predicted by the zero banana width models. In this paper, detailed comparisons between the calculated ICRF power deposition profiles and experimentally measured quantities will be presented which indicate that significant deviations from the zero banana width models occur even for modest power levels (P rf ∼ 6 MW) in the TFTR experiments

Effect of yttrium on the oxide scale adherence of pre-oxidized silicon-containing heat-resistant alloy

International Nuclear Information System (INIS)

Yan Jingbo; Gao Yimin; Shen Yudi; Yang Fang; Yi Dawei; Ye Zhaozhong; Liang Long; Du Yingqian

2011-01-01

Highlights: → AE experiment shows yttrium has a beneficial effect on the pre-oxidized HP40 alloy. → Yttrium facilitates the formation of internal oxide after 10 h of oxidation. → Internal oxide changes the rupture behaviour of the oxide scale. → Twins form in the internal oxide and improve the binding strength of the scale. - Abstract: This paper investigates the effect of the rare earth element yttrium on the rupture behaviour of the oxide scale on the silicon-containing heat-resistant alloy during cooling. After 10 h of oxidation, yttrium is found to facilitate the formation of internal oxides (silica) at the scale-matrix interface. Due to the twinning observed by scanning transmission electron microscopy (STEM) in silica, the critical strain value for the scale failure can be dramatically improved, and the formation of cracks at the scale-matrix interface is inhibited.

Thermal striping heat transfer measurements in sodium AKB experiments

International Nuclear Information System (INIS)

Sheriff, N.; Sephton, K.P.; Gleave, C.

1988-01-01

Temperature fluctuations are produced in the sodium flow of fast reactors where hot and cold flow streams mix. A sodium experiment mounted on the Interatom facility AKB has been used to measure the heat transfer conditions in a flow stream with typical temperature fluctuations. The measurements were made at locations near to the leading edge of a plate, where in practice the most severe conditions are expected. With tests carried out over a wide range of flows good correlations of the heat transfer data with flow have been obtained. A simple theoretical model is proposed to explain the magnitude of the measured heat transfer coefficients, and the use of reasonable assumptions in the model produce good agreement with the experimental measurements

Efficacy of heat treatment for disinfestation of concrete grain silos

Science.gov (United States)

Field experiments were conducted in 2007 and 2008 to evaluate heat treatment for disinfestations of empty concrete elevator silos. A Mobile Heat Treatment Unit was used to introduce heat into silos to attain target conditions of 50°C for at least 6 h. Ventilated plastic containers with a capacity of...

Natural convection heat transfer experiments of horizontal plates with fin arrays

Energy Technology Data Exchange (ETDEWEB)

Moon, Je Young; Chung, Bum Jin [Jeju National University 102 Jejudaehakno, Jeju (Korea, Republic of)

2012-10-15

Core melt in a severe accident condition, forms a molten pool in the reactor vessel lower head. The molten pool is divided by a metallic pool (top) and an oxide pool (bottom) by the density difference. The crust between the metallic layer and the oxide pool may be formed by solidification of the molten metallic materials. So the surface of the crust is formed irregularly. Experiments were performed to investigate the irregular crust as a preparatory study before an in-depth severe accident study. The natural convection heat transfer were investigated experimentally varying the height and spacing of fins, top plate of different kinds and the plate separation distance with/without the side walls. In order to simulate irregular crust surface condition, the finned plates was used. Using the analogy concept, heat transfer experiments were replaced by mass transfer experiments. A cupric acid.copper sulfate (H{sup 2S}O{sup 4-}CuSO{sup 4)} electroplating system was adopted as the mass transfer system and the electric currents were measured rather than the heat transfer rates.

Versatile Desktop Experiment Module (DEMo) on Heat Transfer

Science.gov (United States)

Minerick, Adrienne R.

2010-01-01

This paper outlines a new Desktop Experiment Module (DEMo) engineered for a chemical engineering junior-level Heat Transfer course. This new DEMo learning tool is versatile, fairly inexpensive, and portable such that it can be positioned on student desks throughout a classroom. The DEMo system can illustrate conduction of various materials,…

Decay heat experiment and validation of calculation code systems for fusion reactor

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki

1999-10-01

Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)

Decay heat experiment and validation of calculation code systems for fusion reactor

International Nuclear Information System (INIS)

Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki

1999-10-01

Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)

Recent TMX-U central cell heating and fueling experiments

International Nuclear Information System (INIS)

Hooper, E.B. Jr.; Barter, J.; Dimonte, G.; Falabella, S.; Molvik, A.W.; Pincosy, P.; Turner, W.C.

1986-01-01

Recent experiments have begun to test new methods of heating and fueling of the TMX-U central cell plasma. Heating is with ICRH and 2kV neutral beams. Fueling is by the 2kV beams and by gas puffing. The ICRH system used for fundamental-frequency slow-wave heating consists of two double half-turn antennas, with one on each side of the central cell midplane at mirror ratios of 1:3 and 1:5. Gas fueling is between these two antennas to ensure that recently ionized particles pass through an ICRH resonance before entering the thermal barrier and cells. In recent gas-fed experiments with 100 to 200kW power on each antenna, the end loss temperature was measured to increase from 30eV to above 150eV with perpendicular (cc) temperatures of >500eV. The TMX-U central cell has been equipped with 10 low energy neutral-beam injectors (LENI). These beams are designed to operate at 2kV (net) accel-voltage and deliver 17 atom amperes each to the TMX-U plasma. This low energy was selected to improve trapping (relative to higher energy) on the initial ICRH heated plasma (2X10/sup 12/ cm/sup -3/). At 2keV the beams are predicted to be capable of building up and fueling to 10/sup 13/ cm/sup -3/ density, with ion-ion scattering providing a warm, isotropic ion component in the central cell

Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions

International Nuclear Information System (INIS)

Anderson, M.H.; Corradini, M.L.

1998-01-01

The new AP600 reactor designed by Westinghouse uses a passive safety system relying on heat removal by condensation to keep the containment within the design limits of pressure and temperature. Even though some research has been done so far in this regard, there are some uncertainties concerning the behavior of the system under postulated accident conditions. In this paper, steam condensation onto the internal surfaces of the AP600 containment walls has been investigated in two scaled vessels with similar aspect ratios to the actual AP600. The heat transfer degradation in the presence of noncondensable gas has been analyzed for different noncondensable mixtures of air and helium (hydrogen simulant). Molar fractions of noncondensables/steam ranged from (0.4-4.0) and helium concentrations in the noncondensable mixture were 0-50% by volume. In addition, the effects of the bulk temperatures, the mass fraction of noncondensable/steam, the cold wall surface temperature, the pressure, noncondensable composition, and the inclination of the condensing surface were studied. It was found that the heat transfer coefficients ranged from 50 to 800 J s -1 K -1 m -2 with the highest for high wall temperatures at high pressure and low noncondensable molar fractions. The effect of a light gas (helium) in the noncondensable mixture were found to be negligible for concentrations less than approximately 35 molar percent but could result in stratification at higher concentrations. The complete study gives a large and relatively complete data base on condensation within a scaled AP600 containment structure, providing an invaluable set of data against which to validate models. In addition, specific areas requiring further investigation are summarized. (orig.)

Atmospheric stability inside containments with a heated layer of liquid on the floor

Energy Technology Data Exchange (ETDEWEB)

Vate, J.F. van de [Netherlands Energy Research Foundation, Petten (Netherlands)

1977-01-01

The study of atmospheric stability inside containments with a heated layer of liquid comprised derivation of the boundary condition for stable atmospheric stratifications and the experimental validation of the boundary condition for stable atmospheric stratification. This report includes description of the model for stirred aerosol deposition and the calculation results for maximum aerodynamic diameter of a confined aerosol remaining just well-stirred.

Recent advances in severe accident technology - direct containment heating in advanced light water reactors

International Nuclear Information System (INIS)

Fontana, M.H.

1993-01-01

The issues affecting high-pressure melt ejection (HPME) and the consequential containment pressurization from direct containment heating (DCH), as they affect advanced light water reactors (ALWRs), specifically advanced pressurized water reactors (APWRs), were reviewed by the U.S. Department of Energy Advanced Reactor Severe Accident Program (ARSAP). Recommendations from ARSAP regarding the design of APWRs to minimize DCH are embodied within the Electric Power Research Institute ALWR Utility Requirements Document, which specifies (a) a large, strong containment; (b) an in-containment refueling water storage tank; (c) a reactor cavity configuration that minimizes energy transport to the containment atmosphere; and (d) a reactor coolant system depressurization system. Experimental and analytical efforts, which have focused on current-generation plants, and analyses for APWRs were reviewed. Although DCH is a subject of continuous research and considerable uncertainties remain, it is the judgment of the ARSAP that reactors complying with the recommended design requirements would have a low probability of early containment failure due to HPME and DCH

ICRF heating on the burning plasma experiment (BPX)

International Nuclear Information System (INIS)

Batchelor, D.B.; Carter, M.D.; Goulding, R.H.; Hoffman, D.J.; Jaeger, E.F.; Ryan, P.M.; Swain, D.W.; Tolliver, J.S.; Yugo, J.J.; Goldston, R.J.; Hosea, J.C.; Kaye, S.M.; Phillips, C.K.; Wilson, J.R.; Mau, T.K.

1991-01-01

RF power in the ion cyclotron range of frequencies (ICRF) has been chosen as the primary heating technique for BPX. This decision is based on the wide success of ICRF heating in existing experiments (JET, TFTR, JT-60), the capability of ion cyclotron waves to penetrate the high-density plasmas of BPX, the ability to concentrate ICRF power deposition near the plasma center, and the ready availability of high-power sources at the appropriate frequency. The primary task of the ICRF system is to heat the plasma to ignition. However, other important roles are envisaged; these include the stabilization of sawteeth, preheating of the plasma during current ramp-up, and possible control of the plasma current profile by means of fast-wave current drive. We give a brief overview of the RF system, describe the operating scenarios planned for BPX, and discuss some of the antenna design issues for BPX. 4 refs., 3 figs

Simulations and experiments of laminar heat transfer for Therminol heat transfer fluids in a rifled tube

International Nuclear Information System (INIS)

Xu, Weiguo; Ren, Depeng; Ye, Qing; Liu, Guodong; Lu, Huilin; Wang, Shuai

2016-01-01

Graphical abstract: Predicted laminar Nusselt number using regression correlation of Therminol-55 heat transfer fluid is in agreement with experiments in the rifled tube. - Highlights: • Heat transfer coefficient and friction factor are measured and predicted in the rifled tube. • Correlations for Nusselt number and friction factor are proposed. • The roughness height of 0.425 mm in transition SST model is suggested as an input parameter. • k–kl–ω transition and transition SST models are recommended for laminar–turbulent transition. • Thermal enhancement factor and synergy angle are predicted in the rifled tube. - Abstract: Simulations and experiments of flow and heat transfer behavior of Therminol-55 heat transfer fluid have been conducted in a horizontal rifled tube with outer diameter and inner diameter 25.0 and 20.0 mm, pitch and rib height of 12.0 and 1.0 mm, respectively. Numerical simulations of three-dimensional flow behavior of Therminol-55 heat transfer fluid are carried out using FLUENT code in the rifled tube. Experimental results show that the heat transfer and thermal performance of Therminol-55 heat transfer fluid in the rifled tube are considerably improved compared to those of the smooth tube. The Nusselt number increases with the increase of Reynolds number, and is from 3.5 to 5.1 times over the smooth tube. Also, the pressure drop results reveal that the average friction factor of the ribbed tube is in a range of 2.2 and 4.2 times over the smooth tube. Predictive Nusselt number and friction factor correlations have been presented. The numerical results show that the laminar flow model is valid only at lower Reynolds number in the developed laminar flow of rifled tube. The k–kl–ω transition model and transition SST model with roughness of 0.425 mm are recommended for the predictions of transition process from laminar to turbulent flow in the rifled tube.

Microstructure and heat resistance of Mg-Al-Zn alloys containing metastable phase

International Nuclear Information System (INIS)

Kim, Jeong-Min; Park, Bong-Koo; Jun, Joong-Hwan; Shin, Keesam; Kim, Ki-Tae; Jung, Woon-Jae

2007-01-01

In this research microstructural studies have been made on cast specimens of AZ91 base alloys containing various amounts of Zn. As the amount of Zn addition increased up to 2%, any new Zn-containing phase did not appear while the Zn content in Mg 17 Al 12 phase continuously increased. A quasi-crystalline phase started to form at Mg 17 Al 12 phase when the added Zn content was about 3 wt.%. The tensile strength and elongation of the alloys at 175 deg. C were observed to increase significantly with increasing Zn content. The quasi-crystalline phase was found to be stable up to 300 deg. C, based on scanning electron microscopy examinations of the specimens heated at different temperatures for 24 h

Disruption simulation experiment using high-frequency rastering electron beam as the heat source

International Nuclear Information System (INIS)

Yamazaki, S.; Seki, M.

1987-01-01

The disruption is a serious event which possibly reduces the lifetime of plasm interactive components, so the effects of the resulting high heat flux on the wall materials must be clearly identified. The authors performed disruption simulation experiments to investigate melting, evaporation, and crack initiation behaviors using an electron beam facility as the heat source. The facility was improved with a high-frequency beam rastering system which provided spatially and temporally uniform heat flux on wider test surfaces. Along with the experiments, thermal and mechanical analyses were also performed. A two-dimensional disruption thermal analysis code (DREAM) was developed for the analyses

HTGR nuclear heat source component design and experience

International Nuclear Information System (INIS)

Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

1982-05-01

The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

EIR, KfK joint heat transfer experiment on a single rod, roughened with trapezoidal rounded ribs and cooled by various gases

International Nuclear Information System (INIS)

Hudina, M.; Huggenberger, M.; Dalle Donne, M.; Meyer, L.; Rehme, K.

1978-10-01

Heat transfer and pressure drop experiments with an identical fuel rod simulator have been performed at the two Research Establishments in Wuerenlingen and in Karlsruhe. The rod was artificially roughened with 'two-dimensional' ribs of trapezoidal shape and with rounded edges. The experiments at EIR were performed with CO 2 and the rough rod was contained in a smooth tube and centered by special spacers. The experiments at KfK were performed using helium and nitrogen in the same test section (rod and outer smooth tube with spacers) and air with the rod mounted in another outer smooth tube and with spacers located farther away from the measuring positions. The global measured friction factors and Stanton numbers for different gases agree reasonably well. The differences between the roughness parameters R and G are larger. The possible reasons for the uncertainties in the reduction of these parameters are discussed. It is recommended to perform further experiments with helium in a test section with spacers far apart, to investigate the effect of the temperature ratio on heat transfer and to check if the unfavourable data obtained by the experiment with air are too pessimistic for GCFR application. (orig.) [de

Heat Transfer Experiment with Supercritical CO2 Flowing Upward in a Circular Tube

International Nuclear Information System (INIS)

Kim, Hyung Rae; Kim, Hwan Yeol; Song, Jin Ho; Kim, Hee Dong; Bae, Yoon Yeong

2005-01-01

SCWR (SuperCritical Water-cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project, which aims at the development of new reactors with enhanced economy and safety. Heat transfer experiments under supercritical conditions are required in relevant geometries for the proper prediction of thermo-hydraulic phenomena in a reactor core. A heat transfer test loop, named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt generation), has been constructed in KAERI. The loop uses carbon dioxide as a surrogate fluid for water since the critical pressure and temperature of CO 2 are much lower those of water. As a first stage of heat transfer experiments, a single tube test is being performed in the test loop. Controlled parameters for the tests are operating pressure, mass flux, and heat flux. Wall temperatures are measured along the tube. Experimental data are compared with existing correlations

Standard Practice for Laboratory Screening of Metallic Containment Materials for Use With Liquids in Solar Heating and Cooling Systems

CERN Document Server

American Society for Testing and Materials. Philadelphia

1980-01-01

1.1 This practice covers several laboratory test procedures for evaluating corrosion performance of metallic containment materials under conditions similar to those that may occur in solar heating and cooling systems. All test results relate to the performance of the metallic containment material only as a part of a metal/fluid pair. Performance in these laboratory test procedures, taken by itself, does not necessarily constitute an adequate basis for acceptance or rejection of a particular metal/fluid pair in solar heating and cooling systems, either in general or in a particular design. This practice is not intended to preclude the use of other screening tests, particularly when those tests are designed to more closely simulate field service conditions. 1.2 This practice describes apparatus and procedures for several tests, any one or more of which may be used to evaluate the deterioration of the metallic containment material in a metal/fluid pair. The procedures are designed to permit simulation, heating...

The buffer/container experiment design and construction report

Energy Technology Data Exchange (ETDEWEB)

Chandler, N.A.; Wan, A.W.L.; Roach, P.J

1998-03-01

The Buffer/Container Experiment was a full-scale in situ experiment, installed at a depth of 240 m in granitic rock at AECL's Underground Research Laboratory (URL). The experiment was designed to examine the performance of a compacted sand-bentonite buffer material under the influences of elevated temperature and in situ moisture conditions. Buffer material was compacted in situ into a 5-m-deep, 1.24-m-diameter borehole drilled into the floor of an excavation. A 2.3-m long heater, representative of a nuclear fuel waste container, was placed within the buffer, and instrumentation was installed to monitor changes in buffer moisture conditions, temperature and stress. The experiment was sealed at the top of the borehole and restrained against vertical displacement. Instrumentation in the rock monitored pore pressures, temperatures and rock displacement. The heater was operated at a constant power of 1200 W, which provided a heater skin temperature of approximately 85 degrees C. Experiment construction and installation required two years, followed by two and a half years of heater operation and two years of monitoring the rock conditions during cooling. The construction phase of the experiment included the design, construction and testing of a segmental heater and controller, geological and hydrogeological characterization of the rock, excavation of the experiment room, drilling of the emplacement borehole using high pressure water, mixing and in situ compaction of buffer material, installation of instrumentation in the rock, buffer and on the heater, and the construction of concrete curb and steel vertical restraint system at the top of emplacement borehole. Upon completion of the experiment, decommissioning sampling equipment was designed and constructed and sampling methods were developed which allowed approximately 2000 samples of buffer material to be taken over a 12-day period. Quality assurance procedures were developed for all aspects of experiment

The buffer/container experiment design and construction report

International Nuclear Information System (INIS)

Chandler, N.A.; Wan, A.W.L.; Roach, P.J.

1998-03-01

The Buffer/Container Experiment was a full-scale in situ experiment, installed at a depth of 240 m in granitic rock at AECL's Underground Research Laboratory (URL). The experiment was designed to examine the performance of a compacted sand-bentonite buffer material under the influences of elevated temperature and in situ moisture conditions. Buffer material was compacted in situ into a 5-m-deep, 1.24-m-diameter borehole drilled into the floor of an excavation. A 2.3-m long heater, representative of a nuclear fuel waste container, was placed within the buffer, and instrumentation was installed to monitor changes in buffer moisture conditions, temperature and stress. The experiment was sealed at the top of the borehole and restrained against vertical displacement. Instrumentation in the rock monitored pore pressures, temperatures and rock displacement. The heater was operated at a constant power of 1200 W, which provided a heater skin temperature of approximately 85 degrees C. Experiment construction and installation required two years, followed by two and a half years of heater operation and two years of monitoring the rock conditions during cooling. The construction phase of the experiment included the design, construction and testing of a segmental heater and controller, geological and hydrogeological characterization of the rock, excavation of the experiment room, drilling of the emplacement borehole using high pressure water, mixing and in situ compaction of buffer material, installation of instrumentation in the rock, buffer and on the heater, and the construction of concrete curb and steel vertical restraint system at the top of emplacement borehole. Upon completion of the experiment, decommissioning sampling equipment was designed and constructed and sampling methods were developed which allowed approximately 2000 samples of buffer material to be taken over a 12-day period. Quality assurance procedures were developed for all aspects of experiment construction

Nuclear reactor container

International Nuclear Information System (INIS)

Ishiyama, Takenori.

1989-01-01

This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

Fast wave heating experiments in the ion cyclotron range of frequencies on ATF

Energy Technology Data Exchange (ETDEWEB)

Kwon, M; Shepard, T D; Goulding, R H [Oak Ridge National Lab., TN (United States); and others

1992-07-01

Fast wave heating experiments in the ion cyclotron range of frequencies (ICRF) were performed on target plasmas produced by 350 kW of electron cyclotron heating at 53 GHz and also by neutral beam injection in the Advanced Toroidal Facility (ATF). Various heating regimes were investigated in the frequency range between 9.2 MHz and 28.8 MHz with magnetic fields of 0.95 T and 1.9 T on axis. The nominal pulse lengths of up to 200 kW RF power were in the range between 100 and 400 ms. Data from spectroscopy, loading measurements, and edge RF and Langmuir probes were used to characterize the RF induced effects on the ATF plasma. In the hydrogen minority regime at low plasma density, large suprathermal ion tails were observed with a neutral particle analyser. At high density (n-bar{sub e} {>=} 5.0 x 10{sup 13} cm{sup -3}) substantial increases in antenna loading were observed, but ICRF power was insufficient to produce definitive heating results. A two-dimensional RF heating code, ORION, and a Fokker-Planck code, RFTRANS, were used to simulate these experiments. A simulation of future high power, higher density experiments in ATF indicates improved bulk heating results due to the improved loading and more efficient thermalization of the minority tail. (author). 29 refs, 16 figs, 3 tabs.

Storage vessel for radiation contaminated container

International Nuclear Information System (INIS)

Sakatani, Tadatsugu.

1996-01-01

In a storage vessel of the present invention, a plurality of radiation contaminated material containing bodies are vertically stacked in a cell chamber. Then, the storage vessel comprises a containing tube for containing a plurality of the containing bodies, cooling coils wound around the containing tube, a cooling medium circulating system connected to the cooling coils and circulating cooling medium, and a heat exchanger interposed to the cooling medium circulating system for removing heat of the cooling medium. Heat of the radioactive material containing bodies is transferred to cooling air and cooling coils by way of the container tube, thereby cooling the containing bodies. By the operation of circulating pumps in a cooling medium circulation system, the cooling medium circulates through a circulation channel comprising a cooling medium transfer pipes, cooling medium branching tubes, cooling coils and the heat exchanger, then heat of the cooling medium is transferred to a heat utilizing system by way of the heat exchanger to attain effective utilization of the heat. In this case, heat can be taken out stably even when the storage amount fluctuates and heat releasing amount is reduced, and improvement of heat transfer promotes the cooling of the containing bodies, which enables minimization of the size of the storage vessel. (T.M.)

Energy confinement scaling in tokamaks: some implications of recent experiments with ohmic and strong auxiliary heating

International Nuclear Information System (INIS)

Goldston, R.J.

1984-02-01

Recent results from confinement scaling experiments on tokamaks with ohmic and strong auxiliary heating are reviewed. An attempt is made to draw these results together into a low-density ohmic confinement scaling law, and a scaling law for confinement with auxiliary heating. The auxiliary heating confinement law may also serve to explain the saturation in tau/sub E/ vs anti n/sub e/ observed in some ohmic heating density scaling experiments

Modeling of N2 and O optical emissions for ionosphere HF powerful heating experiments

Science.gov (United States)

Sergienko, T.; Gustavsson, B.

Analyses of experiments of F region ionosphere modification by HF powerful radio waves show that optical observations are very useful tools for diagnosing of the interaction of the probing radio wave with the ionospheric plasma Hitherto the emissions usually measured in the heating experiment have been the 630 0 nm and the 557 7 nm lines of atomic oxygen Other emissions for instance O 844 8 nm and N2 427 8 nm have been measured episodically in only a few experiments although the very rich optical spectrum of molecular nitrogen potentially involves important information about ionospheric plasma in the heated region This study addresses the modeling of optical emissions from the O and the N2 triplet states first positive second positive Vegard-Kaplan infrared afterglow and Wu-Benesch band systems excited under a condition of the ionosphere heating experiment The auroral triplet state population distribution model was modified for the ionosphere heating conditions by using the different electron distribution functions suggested by Mishin et al 2000 2003 and Gustavsson at al 2004 2005 Modeling results are discussed from the point of view of efficiency of measurements of the N2 emissions in future experiments

Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

Energy Technology Data Exchange (ETDEWEB)

Fritsching, G.; Wolf, G. [Internationale Atomreaktorbau G.m.b.H. (INTERATOM), Bergisch Gladbach (Germany, F.R.)

1981-01-15

The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen.

Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

International Nuclear Information System (INIS)

Fritsching, G.; Wolf, G.

1981-01-01

The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen

Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

International Nuclear Information System (INIS)

Zhao, Hangbin; Yan, Changqi; Sun, Licheng; Zhao, Kaibin; Fa, Dan

2015-01-01

As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

Simulation of the VISTA SG heat transfer experiment using MIDAS/SMR

International Nuclear Information System (INIS)

Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

2011-01-01

As the SMART plant was designed with the helical type tubes in the steam generators, the heat transfer model in that geometry has been implemented in the TASS/SMR-S code and used for the safety analysis. The same correlation was implemented in the MIDAS/SMR, which is being used for the severe accident analyses, to model heat transfer at the steam generators. In this study, the VISTA SG experiment with the helical steam generator tube was simulated with MIDAS/SMR to compare the heat transfer rates through the helical tube

Review of ICRF antenna development and heating experiments up to advanced experiment I, 1989 on the JT-60 tokamak

International Nuclear Information System (INIS)

Fujii, Tsuneyuki

1992-03-01

Two main subjects of ion cyclotron range of frequencies (ICRF) heating on JT-60 are described in this paper from development phase of the JT-60 ICRF heating system up to advanced experiment I, 1989. One is antenna design and development for the high power JT-60 ICRF heating system (6 MW for 10 s at a frequency range of 108 - 132 MHz). The other is the experimental investigation of characteristics of second harmonic ICRF heating in a large tokamak. (J.P.N.)

Fundamental study on the melting process of crushed ice in a heat storage container; Chikunetsu sonai ni takuwaeta saihyo no yukai ni kansuru kisoteki kenkyu

Energy Technology Data Exchange (ETDEWEB)

Yanadori, M; Kobori, H [Hitachi, Ltd., Tokyo (Japan); Tsubota, Y [Tokyo Electric Power Co. Inc., Tokyo (Japan)

1998-03-25

This report deals with heat transfer in the melting process of crushed ice filling in a ice/water heat storage container. Volumetric heat transfer rate and melting end-time are measured when rectangular-type, small-stone-type and particle-type ice in the container are melted by circulation hot water. Melting end-time of small-stone-type ice is the shortest and that of particle-type ice is the latest. Volumetric heat transfer rate of small-stone-type ice and rectangular-type ice is larger than that of particle-type ice. The flow rate of circulation hot water throwing in container through a inlet pipe influences remarkably on heat transfer rate. 4 refs., 10 figs.

Required momentum, heat, and mass transport experiments for liquid-metal blankets

International Nuclear Information System (INIS)

Tillack, M.S.; Sze, D.K.; Abdou, M.A.

1986-01-01

Through the effects on fluid flow, many aspects of blanket behavior are affected by magnetohydrodynamic (MHD) effects, including pressure drop, heat transfer, mass transfer, and structural behavior. In this paper, a set of experiments is examined that could be performed in order to reduce the uncertainties in the highly related set of issues dealing with momentum, heat, and mass transport under the influence of a strong magnetic field (i.e., magnetic transport phenomena). By improving our basic understanding and by providing direct experimental data on blanket behavior, these experiments will lead to improved designs and an accurate assessment of the attractiveness of liquid-metal blankets

Experiment of forced convection heat transfer using microencapsulated phase-change-material slurries

International Nuclear Information System (INIS)

Kubo, Shinji; Akino, Norio; Tanaka, Amane; Nagashima, Akira.

1997-01-01

The present study describes an experiment on forced convective heat transfer using a water slurry of Microencapsulated Phase-change-material. A normal paraffin hydrocarbon is microencapsulated by melamine resin, melting point of 28.1degC. The heat transfer coefficient and pressure drop in a circular tube were evaluated. The heat transfer coefficient using the slurry in case with and without phase change were compared to in case of using pure water. (author)

New method of processing heat treatment experiments with numerical simulation support

Science.gov (United States)

Kik, T.; Moravec, J.; Novakova, I.

2017-08-01

In this work, benefits of combining modern software for numerical simulations of welding processes with laboratory research was described. Proposed new method of processing heat treatment experiments leading to obtaining relevant input data for numerical simulations of heat treatment of large parts was presented. It is now possible, by using experiments on small tested samples, to simulate cooling conditions comparable with cooling of bigger parts. Results from this method of testing makes current boundary conditions during real cooling process more accurate, but also can be used for improvement of software databases and optimization of a computational models. The point is to precise the computation of temperature fields for large scale hardening parts based on new method of temperature dependence determination of the heat transfer coefficient into hardening media for the particular material, defined maximal thickness of processed part and cooling conditions. In the paper we will also present an example of the comparison standard and modified (according to newly suggested methodology) heat transfer coefficient data’s and theirs influence on the simulation results. It shows how even the small changes influence mainly on distribution of temperature, metallurgical phases, hardness and stresses distribution. By this experiment it is also possible to obtain not only input data and data enabling optimization of computational model but at the same time also verification data. The greatest advantage of described method is independence of used cooling media type.

Study on the application of thickened welds without post weld heat treatment for containment vessels

International Nuclear Information System (INIS)

Takeuchi, T.; Fukaya, T.; Sato, M.; Takano, G.

1978-01-01

As material for containment vessels, SGV49 steel plates are mainly used. However, those used for this purpose are limited in thickness to smaller than 38 mm. This is because the present standard requires welds thicker than 38 mm to be subjected to post weld heat treatment but operation on the site is practically difficult. In the case of 3-loop containment vessels of pressurized water type reactors, use of 38 mm SGV49 brings an increase in their height and this is disadvantageous from a seismic viewpoint. Therefore, use of 45 mm-thick steel material has become necessary in order to increase design internal pressure and reduce the height of the vessels. To investigate the propriety of the use of 45 mm-thick SGV49 for this purpose without post weld heat treatment we investigated the basic performances of base metal and welded joints. We also conducted large-scale embrittlement fracture tests (CT test, deep notch test, wide plate tensile test and ESSO test) in order to examine whether welds not subjected to post weld heat treatment are safe against embrittlement fracture under the operating conditions of the vessels. The results proved that the welds of SGV49 steel plates are safe enough under the operating conditions. (author)

Reactor container

International Nuclear Information System (INIS)

Shibata, Satoru; Kawashima, Hiroaki

1984-01-01

Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

Preliminary results of statistical dynamic experiments on a heat exchanger

International Nuclear Information System (INIS)

Corran, E.R.; Cummins, J.D.

1962-10-01

The inherent noise signals present in a heat exchanger have been recorded and analysed in order to determine some of the statistical dynamic characteristics of the heat exchanger. These preliminary results show that the primary side temperature frequency response may be determined by analysing the inherent noise. The secondary side temperature frequency response and cross coupled temperature frequency responses between primary and secondary are poorly determined because of the presence of a non-stationary noise source in the secondary circuit of this heat exchanger. This may be overcome by correlating the dependent variables with an externally applied noise signal. Some preliminary experiments with an externally applied random telegraph type of signal are reported. (author)

Experiments Demonstrate Geothermal Heating Process

Science.gov (United States)

Roman, Harry T.

2012-01-01

When engineers design heat-pump-based geothermal heating systems for homes and other buildings, they can use coil loops buried around the perimeter of the structure to gather low-grade heat from the earth. As an alternative approach, they can drill well casings and store the summer's heat deep in the earth, then bring it back in the winter to warm…

Heat tracer test in an alluvial aquifer: Field experiment and inverse modelling

Science.gov (United States)

Klepikova, Maria; Wildemeersch, Samuel; Hermans, Thomas; Jamin, Pierre; Orban, Philippe; Nguyen, Frédéric; Brouyère, Serge; Dassargues, Alain

2016-09-01

Using heat as an active tracer for aquifer characterization is a topic of increasing interest. In this study, we investigate the potential of using heat tracer tests for characterization of a shallow alluvial aquifer. A thermal tracer test was conducted in the alluvial aquifer of the Meuse River, Belgium. The tracing experiment consisted in simultaneously injecting heated water and a dye tracer in an injection well and monitoring the evolution of groundwater temperature and tracer concentration in the pumping well and in measurement intervals. To get insights in the 3D characteristics of the heat transport mechanisms, temperature data from a large number of observation wells closely spaced along three transects were used. Temperature breakthrough curves in observation wells are contrasted with what would be expected in an ideal layered aquifer. They reveal strongly unequal lateral and vertical components of the transport mechanisms. The observed complex behavior of the heat plume is explained by the groundwater flow gradient on the site and heterogeneities in the hydraulic conductivity field. Moreover, due to high injection temperatures during the field experiment a temperature-induced fluid density effect on heat transport occurred. By using a flow and heat transport numerical model with variable density coupled with a pilot point approach for inversion of the hydraulic conductivity field, the main preferential flow paths were delineated. The successful application of a field heat tracer test at this site suggests that heat tracer tests is a promising approach to image hydraulic conductivity field. This methodology could be applied in aquifer thermal energy storage (ATES) projects for assessing future efficiency that is strongly linked to the hydraulic conductivity variability in the considered aquifer.

GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

International Nuclear Information System (INIS)

Ramachandran, S.; Krause, M.; Nguyen, T.

2003-01-01

This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

Evaluation of conceptual Heat Exchanger Design for passive containment cooling system of SMART

Energy Technology Data Exchange (ETDEWEB)

Kim, Min-Ki; Hong, Soon Joon [FNC Tech., Yongin (Korea, Republic of); Kim, Young In; Kim, Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

PCCS(Passive containment cooling system) is the passive safety system which ultimately removes the reactor decay heat. Cooling performance of the air-cooled type and water-circulation cooling type of PCCS were analyzed using CAP version 2.21. The analysis results show the water-circulation cooling PCCS is more effective in lowering the peak pressure and temperature in the containment building. However, the air-cooled PCCS is more effective to the long-term cooling. From this study, the efficiency evaluation results for the two PCCS designs are obtained. These results may be applied in the PCCS design improvement. Moreover, these results will be used as a reference for the later PCCS design and analysis.

Experience in the use of wind energy for greenhouse heating

Energy Technology Data Exchange (ETDEWEB)

O' Flaherty, T; Kocsis, K; Petersen, H [eds.

1987-05-01

Study of the appliction of wind energy for greenhouse heating began at Kinsealy Research Centre in 1980 with the installation of a multi-blade 6m diamter wind turbine. This produced electricity which was used to provide root zone warming for a glasshouse tomato crop. The application worked well and the wind turbine is still in operation, although it has been out of service for substantial periods and has required major refurbishment. In July 1985 a new wind turbine was commissioned as an EEC Wind Energy Demonstration Project. This is an 11m diameter grid-connected unit, and the project involves using its output to power a heat pump which in turn supplies heat to a greenhouse. The system is operating well and initial performance results have been obtained during the 1985-'86 heating season. The paper summarises the experience to data with both of these projects.

Specific heat jump at T/sub c/ of proximity effect sandwiches containing nonmagnetic localized states

International Nuclear Information System (INIS)

Maneeratankul, S.; Tang, I.M.

1987-01-01

The decrease in the transition temperature and the jump in the specific heat at T/sub c/ of proximity effect sandwiches containing nonmagnetic Anderson impurities in the normal layer are studied. The effects of the resonant scattering by the impurities are treated in the same manner as that used by Kaiser in his study of the effects of resonant scattering on the properties of bulk superconductors. Numerical calculations of the decrease in T/sub c/ and the jump in the specific heat at T/sub c/ as a function of the thickness of the normal layer are presented

Coupled heat transfer model and experiment study of semitransparent barrier materials in aerothermal environment

Science.gov (United States)

Wang, Da-Lin; Qi, Hong

Semi-transparent materials (such as IR optical windows) are widely used for heat protection or transfer, temperature and image measurement, and safety in energy , space, military, and information technology applications. They are used, for instance, ceramic coatings for thermal barriers of spacecrafts or gas turbine blades, and thermal image observation under extreme or some dangerous environments. In this paper, the coupled conduction and radiation heat transfer model is established to describe temperature distribution of semitransparent thermal barrier medium within the aerothermal environment. In order to investigate this numerical model, one semi-transparent sample with black coating was considered, and photothermal properties were measured. At last, Finite Volume Method (FVM) was used to solve the coupled model, and the temperature responses from the sample surfaces were obtained. In addition, experiment study was also taken into account. In the present experiment, aerodynamic heat flux was simulated by one electrical heater, and two experiment cases were designed in terms of the duration of aerodynamic heating. One case is that the heater irradiates one surface of the sample continually until the other surface temperature up to constant, and the other case is that the heater works only 130 s. The surface temperature responses of these two cases were recorded. Finally, FVM model of the coupling conduction-radiation heat transfer was validated based on the experiment study with relative error less than 5%.

Plasma experiments on staged theta pinch, implosion heating experiment and Scyllac feedback-sector experiment

International Nuclear Information System (INIS)

Bartsch, R.R.; Buchenauer, C.J.; Cantrell, E.L.

1977-01-01

Results of the Los Alamos theta-pinch program in three areas of investigation are summarized: 1) In the Staged Theta Pinch, results are reported on the effects of magnetic field amplitude and time history of plasma formation. 2) In the Implosion Heating Experiment, density, internal-magnetic field and neutron measurements yield a consistent picture of the implosion which agrees with kinetic computations and with a simple dynamic model of the ions and magnetic piston. 3) In the Scyllac Feedback-Sector Experiment, the l=1, 0 equilibrium plasma parameters have been adjusted to accommodate the feedback stabilization system. With a uniform toroidal discharge tube the m=1 instability is feedback-stabilized in the vertical direction, and confinement in the toroidal direction is extended by feedback control. Results with a helical discharge tube are also reported. (author)

Heat Fluxes and Evaporation Measurements by Multi-Function Heat Pulse Probe: a Laboratory Experiment

Science.gov (United States)

Sharma, V.; Ciocca, F.; Hopmans, J. W.; Kamai, T.; Lunati, I.; Parlange, M. B.

2012-04-01

Multi Functional Heat Pulse Probes (MFHPP) are multi-needles probes developed in the last years able to measure temperature, thermal properties such as thermal diffusivity and volumetric heat capacity, from which soil moisture is directly retrieved, and electric conductivity (through a Wenner array). They allow the simultaneous measurement of coupled heat, water and solute transport in porous media, then. The use of only one instrument to estimate different quantities in the same volume and almost at the same time significantly reduces the need to interpolate different measurement types in space and time, increasing the ability to study the interdependencies characterizing the coupled transports, especially of water and heat, and water and solute. A three steps laboratory experiment is realized at EPFL to investigate the effectiveness and reliability of the MFHPP responses in a loamy soil from Conthey, Switzerland. In the first step specific calibration curves of volumetric heat capacity and thermal conductivity as function of known volumetric water content are obtained placing the MFHPP in small samplers filled with the soil homogeneously packed at different saturation degrees. The results are compared with literature values. In the second stage the ability of the MFHPP to measure heat fluxes is tested within a homemade thermally insulated calibration box and results are matched with those by two self-calibrating Heatflux plates (from Huxseflux), placed in the same box. In the last step the MFHPP are used to estimate the cumulative subsurface evaporation inside a small column (30 centimeters height per 8 centimeters inner diameter), placed on a scale, filled with the same loamy soil (homogeneously packed and then saturated) and equipped with a vertical array of four MFHPP inserted close to the surface. The subsurface evaporation is calculated from the difference between the net sensible heat and the net heat storage in the volume scanned by the probes, and the

Heat experiment design to estimate temperature dependent thermal properties

International Nuclear Information System (INIS)

Romanovski, M

2008-01-01

Experimental conditions are studied to optimize transient experiments for estimating temperature dependent thermal conductivity and volumetric heat capacity. A mathematical model of a specimen is the one-dimensional heat equation with boundary conditions of the second kind. Thermal properties are assumed to vary nonlinearly with temperature. Experimental conditions refer to the thermal loading scheme, sampling times and sensor location. A numerical model of experimental configurations is studied to elicit the optimal conditions. The numerical solution of the design problem is formulated on a regularization scheme with a stabilizer minimization without a regularization parameter. An explicit design criterion is used to reveal the optimal sensor location, heating duration and flux magnitude. Results obtained indicate that even the strongly nonlinear experimental design problem admits the aggregation of its solution and has a strictly defined optimal measurement scheme. Additional region of temperature measurements with allowable identification error is revealed.

HECLA experiments on interaction between metallic melt and hematite-containing concrete

Energy Technology Data Exchange (ETDEWEB)

Sevon, Tuomo, E-mail: tuomo.sevon@vtt.f [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland); Kinnunen, Tuomo; Virta, Jouko; Holmstroem, Stefan; Kekki, Tommi; Lindholm, Ilona [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland)

2010-10-15

In a hypothetical severe accident in a nuclear power plant, molten materials may come into contact with concrete, causing concrete ablation. In five HECLA experiments the interaction between metallic melt and concrete was investigated by pouring molten stainless steel at almost 1800 {sup o}C into cylindrical concrete crucibles. The tests were transient, i.e. no decay heat simulation was used. The main objective was to test the behavior of the FeSi concrete, containing hematite (Fe{sub 2}O{sub 3}) and siliceous aggregates. This special concrete type is used as a sacrificial layer in the Olkiluoto 3 EPR reactor pit, and very scarce experimental data is available about its behavior at high temperatures. It is concluded that no clear differences between the ablation of FeSi concrete and ordinary siliceous concrete were observed. The ablation depths were small, 25 mm at maximum. No dramatic effects, such as cracking of large pieces of concrete due to the thermal shock, took place. An important side result of the test series was gaining knowledge of the properties of the special concrete type. Chemical analyses were conducted and mechanical properties were measured.

RELAP5 and SIMMER-III code assessment on CIRCE decay heat removal experiments

International Nuclear Information System (INIS)

Bandini, Giacomino; Polidori, Massimiliano; Meloni, Paride; Tarantino, Mariano; Di Piazza, Ivan

2015-01-01

Highlights: • The CIRCE DHR experiments simulate LOHS+LOF transients in LFR systems. • Decay heat removal by natural circulation through immersed heat exchangers is investigated. • The RELAP5 simulation of DHR experiments is presented. • The SIMMER-III simulation of DHR experiments is presented. • The focus is on the transition from forced to natural convection and stratification in a large pool. - Abstract: In the frame of THINS Project of the 7th Framework EU Program on Nuclear Fission Safety, some experiments were carried out on the large scale LBE-cooled CIRCE facility at the ENEA/Brasimone Research Center to investigate relevant safety aspects associated with the removal of decay heat through heat exchangers (HXs) immersed in the primary circuit of a pool-type lead fast reactor (LFR), under loss of heat sink (LOHS) accidental conditions. The start-up and operation of this decay heat removal (DHR) system relies on natural convection on the primary side and then might be affected by coolant mixing and temperature stratification phenomena occurring in the LBE pool. The main objectives of the CIRCE experimental campaign were to verify the behavior of the DHR system under representative accidental conditions and provide a valuable database for the assessment of both CFD and system codes. The reproduced accidental conditions refer to a station blackout scenario, namely a protected LOHS and loss of flow (LOF) transient. In this paper the results of 1D RELAP5 and 2D SIMMER-III simulations are compared with the experimental data of more representative DHR transients T-4 and T-5 in order to verify the capability of these codes to reproduce both forced and natural convection conditions observed in the primary circuit and the right operation of the DHR system for decay heat removal. Both codes are able to reproduce the stationary conditions and with some uncertainties the transition to natural convection conditions until the end of the transient phase. The trend

Container for spent fuel assembly

International Nuclear Information System (INIS)

Sawai, Takeshi.

1996-01-01

The container of the present invention comprises a container main body having a body portion which can contain spent fuel assemblies and a lid, and heat pipes having an evaporation portion disposed along the outer surface of the spent fuel assemblies to be contained and a condensation portion exposed to the outside of the container main body. Further, the heat pipe is formed spirally at the evaporation portions so as to surround the outer circumference of the spent fuel assemblies, branched into a plurality of portions at the condensation portion, each of the branched portion of the condensation portion being exposed to the outside of the container main body, and is tightly in contact with the periphery of the slit portions disposed to the container main body. Then, since released after heat is transferred to the outside of the container main body from the evaporation portion of the heat pipe along the outer surface of the spent fuel assemblies by way of the condensation portion of the heat pipes exposed to the outside of the container main body, the efficiency of the heat transfer is extremely improved to enhance the effect of removing heat of spent fuel assemblies. Further, cooling effect is enhanced by the spiral form of the evaporation portion and the branched condensation portion. (N.H.)

Heat Transfer Experiment with Supercritical CO{sub 2} Flowing Upward in a Circular Tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyung Rae; Kim, Hwan Yeol; Song, Jin Ho; Kim, Hee Dong; Bae, Yoon Yeong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

SCWR (SuperCritical Water-cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project, which aims at the development of new reactors with enhanced economy and safety. Heat transfer experiments under supercritical conditions are required in relevant geometries for the proper prediction of thermo-hydraulic phenomena in a reactor core. A heat transfer test loop, named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt generation), has been constructed in KAERI. The loop uses carbon dioxide as a surrogate fluid for water since the critical pressure and temperature of CO{sub 2} are much lower those of water. As a first stage of heat transfer experiments, a single tube test is being performed in the test loop. Controlled parameters for the tests are operating pressure, mass flux, and heat flux. Wall temperatures are measured along the tube. Experimental data are compared with existing correlations.

High power plasma heating experiments on the Proto-MPEX facility

Science.gov (United States)

Bigelow, T. S.; Beers, C. J.; Biewer, T. M.; Caneses, J. F.; Caughman, J. B. O.; Diem, S. J.; Goulding, R. H.; Green, D. L.; Kafle, N.; Rapp, J.; Showers, M. A.

2017-10-01

Work is underway to maximize the power delivered to the plasma that is available from heating sources installed on the Prototype Materials Plasma Exposure eXperiment (Proto-MPEX) at ORNL. Proto-MPEX is a linear device that has a >100 kW, 13.56 MHz helicon plasma generator available and is intended for material sample exposure to plasmas. Additional plasma heating systems include a 10 kW 18 GHz electron cyclotron heating (ECH) system, a 25 kW 8 MHz ion cyclotron heating ICH system, and a 200 kW 28 GHz electron Bernstein wave (EBW) and ECH system. Most of the heating systems have relatively good power transmission efficiency, however, the 28 GHz EBW system has a lower efficiency owing to stringent requirements on the microwave launch characteristics for EBW coupling combined with the lower output mode purity of the early-model gyrotron in use and its compact mode converter system. A goal for the Proto-MPEX is to have a combined heating power of 200 kW injected into the plasma. Infrared emission diagnostics of the target plate combined with Thomson Scattering, Langmuir probe, and energy analyzer measurements near the target are utilized to characterize the plasmas and coupling efficiency of the heating systems. ORNL is managed by UT-Battelle, LLC, for the U.S. DOE under contract DE-AC-05-00OR22725.

Heat Transfer by Thermo-Capillary Convection. Sounding Rocket COMPERE Experiment SOURCE

Science.gov (United States)

Fuhrmann, Eckart; Dreyer, Michael

2009-08-01

This paper describes the results of a sounding rocket experiment which was partly dedicated to study the heat transfer from a hot wall to a cold liquid with a free surface. Natural or buoyancy-driven convection does not occur in the compensated gravity environment of a ballistic phase. Thermo-capillary convection driven by a temperature gradient along the free surface always occurs if a non-condensable gas is present. This convection increases the heat transfer compared to a pure conductive case. Heat transfer correlations are needed to predict temperature distributions in the tanks of cryogenic upper stages. Future upper stages of the European Ariane V rocket have mission scenarios with multiple ballistic phases. The aims of this paper and of the COMPERE group (French-German research group on propellant behavior in rocket tanks) in general are to provide basic knowledge, correlations and computer models to predict the thermo-fluid behavior of cryogenic propellants for future mission scenarios. Temperature and surface location data from the flight have been compared with numerical calculations to get the heat flux from the wall to the liquid. Since the heat flux measurements along the walls of the transparent test cell were not possible, the analysis of the heat transfer coefficient relies therefore on the numerical modeling which was validated with the flight data. The coincidence between experiment and simulation is fairly good and allows presenting the data in form of a Nusselt number which depends on a characteristic Reynolds number and the Prandtl number. The results are useful for further benchmarking of Computational Fluid Dynamics (CFD) codes such as FLOW-3D and FLUENT, and for the design of future upper stage propellant tanks.

Real time plasma control experiments using the JET auxiliary plasma heating systems as the actuator

International Nuclear Information System (INIS)

Zornig, N.H.

1999-01-01

The role of the Real Time Power Control system (RTPC) in the Joint European Torus (JET) is described in depth. The modes of operation are discussed in detail and a number of successful experiments are described. These experiments prove that RTPC can be used for a wide range of experiments, including: (1) Feedback control of plasma parameters in real time using Ion Cyclotron Resonance Heating (ICRH) or Neutral Beam Heating (NBH) as the actuator in various JET operating regimes. It is demonstrated that in a multi-parameter space it is not sufficient to control one global plasma parameter in order to avoid performance limiting events. (2) Restricting neutron production and subsequent machine activation resulting from high performance pulses. (3) The simulation of α-particle heating effects in a DT-plasma in a D-only plasma. The heating properties of α-particles are simulated using ICRH-power, which is adjusted in real time. The simulation of α-particle heating in JET allows the effects of a change in isotopic mass to be separated from α-particle heating. However, the change in isotopic mass of the plasma ions appears to affect not only the global energy confinement time (τ E ) but also other parameters such as the electron temperature at the plasma edge. This also affects τ E , making it difficult to make a conclusive statement about any isotopic effect. (4) For future JET experiments a scheme has been designed which simulates the behaviour of a fusion reactor experimentally. The design parameters of the International Thermonuclear Experimental Reactor (ITER) are used. In the proposed scheme the most relevant dimensionless plasma parameters are similar in JET and ITER. It is also shown how the amount of heating may be simulated in real time by RTPC using the electron temperature and density as input parameters. The results of two demonstration experiments are presented. (author)

A heat transfer analysis of the CCI experiments 1-3

International Nuclear Information System (INIS)

Sevon, Tuomo

2008-01-01

This paper presents an attempt to evaluate the heat transfer rates and gas release rates in the CCI core-concrete interaction experiments 1-3, performed within the OECD MCCI project. A new method for calculating the heat transfer rates has been developed. It is based on calculating integrals of the concrete enthalpies with the help of piecewise exponential interpolation curves. The new method takes into account heat conduction in the concrete. Compared to traditional methods, the new method gives better results during slow concrete ablation, and its time resolution is significantly better. The gas release rates from the concrete were also calculated. A regression analysis was conducted for the heat transfer coefficients and gas release rates. Three correlations for the bubbling-enhanced heat transfer were developed. For the basemat, a single correlation can be used for both siliceous and limestone/common sand (LCS) concrete types. For the sidewall, two different correlations are needed for the two concrete types. With the same superficial gas velocity, the heat transfer rate to siliceous sidewalls is higher than to LCS sidewalls. This suggests that the reason for the different radial ablation rates of the concrete types observed in the tests is not the lower gas content of siliceous concrete

Core heat transfer experiment for JRR-3 to be upgraded at 20 MWt, 2

International Nuclear Information System (INIS)

Sudo, Yukio; Miyata, Keiichi; Ikawa, Hiromasa; Ohgawara, Masami; Kaminaga, Masanori

1985-09-01

Experiments were carried out to investigate the condition of onset of nucleate boiling (ONB) and the departure from nucleate boiling (DNB) heat flux under forced convection in a vertical rectangular channel, both of which take important roles in the core thermal-hydraulic design of the upgraded JRR-3. This report presents the validity and applicability of the correlations proposed for ONB condition and DNB heat flux, based on the analysis of the experimental results. The upgraded JRR-3 is a low-pressure, low-temperature research reactor and the core heat generation is removed by two cooling modes, one is natural circulation under upflow up to 200 kW and the other is forced circulation under downflow up to 20 MW. Therefore, the difference in heat transfer characteristics between upflow and downflow were investigated in the experiments, which were carried out by using a heated channel properly simulating a subchannel of fuel element because the heat transfer characteristics are considered to be strongly dependent on the configuration of flow channel. (author)

Staged theta pinch experiments

International Nuclear Information System (INIS)

Linford, R.K.; Downing, J.N.; Gribble, R.F.; Jacobson, A.R.; Platts, D.A.; Thomas, K.S.

1976-01-01

Two implosion heating circuits are being experimentally tested. The principal experiment in the program is the 4.5-m-long Staged Theta Pinch (STP). It uses two relatively low energy (50kJ and 100 kJ), high voltage (125 kV) capacitor banks to produce the theta pinch plasma inside the 20 cm i.d. quartz discharge tube. A lower voltage (50 kV), higher energy (750 kJ) capacitor bank is used to contain the plasma and provide a variable amount of adiabatic compression. Because the experiment produces a higher ratio of implosion heating to compressional heating than conventional theta pinches, it should be capable of producing high temperature plasmas with a much larger ratio of plasma radius to discharge tube radius than has been possible in the past. The Resonant Heating Experiment (RHX) in its initial configuration is the same as a 0.9-m-long section of the high voltage part of the STP experiment and all the plasma results here were obtained with the experiment in that configuration. Part of the implosion bank will be removed and a low inductance crowbar added to convert it to the resonant heating configuration. (U.K.)

Experimental investigation of natural convection heat transfer in volumetrically heated spherical segments. Final report

International Nuclear Information System (INIS)

Asfia, F.; Dhir, V.

1998-03-01

One strategy for preventing the failure of lower head of a nuclear reactor vessel is to flood the concrete cavity with subcooled water in accidents in which relocation of core material into the vessel lower head occurs. After the core material relocates into the vessel, a crust of solid material forms on the inner wall of the vessel, however, most of the pool remains molten and natural convection exists in the pool. At present, uncertainty exists with respect to natural convection heat transfer coefficients between the pool of molten core material and the reactor vessel wall. In the present work, experiments were conducted to examine natural convection heat transfer in internally heated partially filled spherical pools with external cooling. In the experiments, Freon-113 contained in a Pyrex bell jar was used as a test liquid. The pool was bounded with a spherical segment at the bottom, and was heated with magnetrons taken from a conventional microwave oven. The vessel was cooled from the outside with natural convection of water or with nucleate boiling of liquid nitrogen

Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

Energy Technology Data Exchange (ETDEWEB)

Bourganel, St

2002-11-01

After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

Electron-cyclotron heating in the Constance 2 mirror experiment

Energy Technology Data Exchange (ETDEWEB)

Mauel, Michael E.

1982-09-01

Electron cyclotron heating of a highly-ionized plasma in mirror geometry is investigated. The experimental diagnosis of the electron energy distribution and the comparison of the results of this diagnosis with a two dimensional, time-dependent Fokker-Planck simulation are accomplished in four steps. (1) First, the power balance of the heated and unheated Constance 2 plasma is analyzed experimentally. It is concluded that the heated electrons escape the mirror at a rate dominated by a combination of the influx of cool electrons from outside the mirror and the increased loss rate of the ions. (2) The microwave parameters at the resonance zones are then calculated by cold-plasma ray tracing. High N/sub parallel/ waves are launched and for these waves, strong first-pass absorption is predicted. The absorption strength is qualitatively checked in the experiment by surrounding the plasma with non-reflecting liners. (3) A simplified quasilinear theory including the effect of N/sub parallel/ is developed to model the electrons. An analytic expression is derived for the RF-induced pump-out of the magnetically-confined warm electrons. Results of the Fokker-Planck simulations show the development of the electron energy distribution for several plasma conditions and verify the scaling of the analytic expression for RF-induced diffusion into the loss cone. (4) Sample x-ray and endloss data are presented, and the overall comparison between the simulation and experiment is discussed. The x-ray signals indicate that, for greater RF power, the hot electrondensity increases more rapidly than its temperature. The time history of the endloss data, illustrating RF-enhancement, suggests the predicted scaling for warm-electron pump-out. Finally, a comparison between the measured and predicted energy distribution shows that the bulk, warm and hot components of the heated Constance 2 electrons are indeed reproduced by the simulation.

Dryout heat flux experiments with deep heterogeneous particle bed

International Nuclear Information System (INIS)

Lindholm, I.; Holmstroem, S.; Miettinen, J.; Lestinen, V.; Hyvaerinen, J.; Pankakoski, P.; Sjoevall, H.

2006-01-01

A test facility has been constructed at Technical Research Centre of Finland (VTT) to simulate as accurately as possible the ex-vessel core particle bed in the conditions of Olkiluoto nuclear power plant. The STYX particle bed reproduces the anticipated depth of the bed and the size range of particles having irregular shape. The bed is immersed in water, creating top flooding conditions, and internally heated by an array of electrical resistance heating elements. Dryout tests have been successfully conducted at 0.1-0.7 MPa pressure for both uniformly mixed and stratified bed geometries. In all tests, including the stratified ones, the dry zone first formed near the bottom of the bed. The measured dryout heat fluxes increased with increasing pressure, from 232 kW/m 2 at near atmospheric pressure to 451 kW/m 2 at 0.7 MPa pressure. The data show some scatter even for the uniform bed. The tests with the stratified bed indicate a clear reduction of critical power due to the presence of a layer of small particles on top of the uniform bed. Comparison of data with various critical power (dryout heat flux) correlations for porous media shows that the most important parameter in the models is the effective particle diameter. Adiabatic debris bed flow resistance measurements were conducted to determine the most representative particle diameter. This diameter is close, but not equal, to the particle number-weighted average diameter of the bed material. With it, uniform bed data can be calculated to within an accuracy of 3-28% using Lipinski's 0-D model. In the stratified bed experiments, it appears that the top layer was partially fluidized, hence the measured critical power was significantly higher than calculated. Future experiments are being planned with denser top layer material to eliminate non-prototypic fluidization

Electron-cyclotron heating in the Constance 2 mirror experiment

International Nuclear Information System (INIS)

Mauel, M.E.

1982-09-01

Electron cyclotron heating of a highly-ionized plasma in mirror geometry is investigated. The experimental diagnosis of the electron energy distribution and the comparison of the results of this diagnosis with a two dimensional, time-dependent Fokker-Planck simulation are accomplished in four steps. (1) First, the power balance of the heated and unheated Constance 2 plasma is analyzed experimentally. It is concluded that the heated electrons escape the mirror at a rate dominated by a combination of the influx of cool electrons from outside the mirror and the increased loss rate of the ions. (2) The microwave parameters at the resonance zones are then calculated by cold-plasma ray tracing. High N/sub parallel/ waves are launched and for these waves, strong first-pass absorption is predicted. The absorption strength is qualitatively checked in the experiment by surrounding the plasma with non-reflecting liners. (3) A simplified quasilinear theory including the effect of N/sub parallel/ is developed to model the electrons. An analytic expression is derived for the RF-induced pump-out of the magnetically-confined warm electrons. Results of the Fokker-Planck simulations show the development of the electron energy distribution for several plasma conditions and verify the scaling of the analytic expression for RF-induced diffusion into the loss cone. (4) Sample x-ray and endloss data are presented, and the overall comparison between the simulation and experiment is discussed. The x-ray signals indicate that, for greater RF power, the hot electrondensity increases more rapidly than its temperature. The time history of the endloss data, illustrating RF-enhancement, suggests the predicted scaling for warm-electron pump-out. Finally, a comparison between the measured and predicted energy distribution shows that the bulk, warm and hot components of the heated Constance 2 electrons are indeed reproduced by the simulation

Numerical analysis of heat transfer of canned liquid foods containing fibers or particles during sterilization

Energy Technology Data Exchange (ETDEWEB)

Wang, Q.Z.; Sakai, N.; Hanzawa, T. [Tokyo Univ. of Fisheries, Tokyo (Japan). Dept. of Food Science and Tech.

2000-10-01

The velocity profile, temperature distribution, and the slowest heating point of a canned liquid food containing fibers or particles were calculated numerically by using fundamental equations that take account of the effect of free convection in the can at an unsteady state under the assumption of imaginary fluid with apparent physical properties. To check these calculated results, the temperature distribution in the can was measured experimentally under the same operating conditions as those of the theoretical analysis. The calculated results agree closely with the experimental ones. Adaptable ranges of present numerical analysis and the positional characteristics of the slowest heating point are shown. (author)

New hybrid nanofluid containing encapsulated paraffin wax and sand nanoparticles in propylene glycol-water mixture: Potential heat transfer fluid for energy management

International Nuclear Information System (INIS)

Manikandan, S.; Rajan, K.S.

2017-01-01

Highlights: • Hybrid nanofluid containing sand nanoparticles & encapsulated paraffin wax prepared. • Specific heat of hybrid nanofluid 9% greater than that of PG-water mixture. • Specific heat & thermal conductivity enhanced at optimum paraffin wax concentration. • Hybrid nanofluid with 1 wt.% paraffin wax & 1 vol% sand nanoparticles best suited. - Abstract: The reduction in specific heat commonly encountered due to the addition of nanoparticles to a heat transfer fluid such as propylene glycol-water mixture, can be overcome by co-dispersing surfactant-encapsulated paraffin wax, leading to formation of a hybrid nanofluid. Experimental investigations have been carried out on the preparation and evaluation of thermophysical properties of a hybrid nanofluid containing pluronic P-123 encapsulated paraffin wax (70–120 nm diameter, 1–5 wt.%) and sand nanoparticles (1 vol%) in propylene glycol-water mixture. The comparison of results of differential scanning calorimetry of pure paraffin wax and encapsulated paraffin wax revealed encapsulation efficiency of 84.4%. The specific heat of hybrid nanofluids monotonously increased with paraffin wax concentration, with 9.1% enhancement in specific heat for hybrid nanofluid containing 5 wt.% paraffin wax, in comparison to propylene glycol-water mixture. There exists an optimum paraffin wax concentration (1 wt.%) for the hybrid nanofluid at which the combination of various thermophysical properties such as specific heat, thermal conductivity and viscosity are favorable for use as heat transfer fluid. Such a hybrid nanofluid can be used as a substitute for propylene glycol-water mixture in solar thermal systems.

Cavity Heating Experiments Supporting Shuttle Columbia Accident Investigation

Science.gov (United States)

Everhart, Joel L.; Berger, Karen T.; Bey, Kim S.; Merski, N. Ronald; Wood, William A.

2011-01-01

The two-color thermographic phosphor method has been used to map the local heating augmentation of scaled idealized cavities at conditions simulating the windward surface of the Shuttle Orbiter Columbia during flight STS-107. Two experiments initiated in support of the Columbia Accident Investigation were conducted in the Langley 20-Inch Mach 6 Tunnel. Generally, the first test series evaluated open (length-to-depth less than 10) rectangular cavity geometries proposed as possible damage scenarios resulting from foam and ice impact during launch at several discrete locations on the vehicle windward surface, though some closed (length-to-depth greater than 13) geometries were briefly examined. The second test series was designed to parametrically evaluate heating augmentation in closed rectangular cavities. The tests were conducted under laminar cavity entry conditions over a range of local boundary layer edge-flow parameters typical of re-entry. Cavity design parameters were developed using laminar computational predictions, while the experimental boundary layer state conditions were inferred from the heating measurements. An analysis of the aeroheating caused by cavities allowed exclusion of non-breeching damage from the possible loss scenarios being considered during the investigation.

Large-scale experiments on aerosol behavior in light water reactor containments

International Nuclear Information System (INIS)

Schock, W.; Bunz, H.; Adams, R.E.; Tobias, M.L.; Rahn, F.J.

1988-01-01

Recently, three large-scale experimental programs were carried out dealing with the behavior of aerosols during core-melt accidents in light water reactors (LWRs). In the Nuclear Safety Pilot Plant (NSPP) program, the principal behaviors of different insoluble aerosols and of mixed aerosols were measured in dry air atmospheres and in condensing steam-air atmospheres contained in a 38-m/sup 3/ steel vessel. The Demonstration of Nuclear Aerosol Behavior (DEMONA) program used a 640-m/sup 3/ concrete containment model to simulate typical accident sequence conditions, and measured the behavior of different insoluble aerosols and mixed aerosols in condensing and transient atmospheric conditions. Part of the LWR Aerosol Containment Experiments (LACE) program was also devoted to aerosol behavior in containment; and 852-m/sup 3/ steel vessel was used, and the aerosols were composed of mixtures of insoluble and soluble species. The results of these experiments provide a suitable data base for validation of aerosol behavior codes. Fundamental insight into details of aerosol behavior in condensing environments has been gained through the results of the NSPP tests. Code comparisons have been and are being performed in the DEMONA and LACE experiments

Simulation of containment atmosphere stratification experiment using local instantaneous description

International Nuclear Information System (INIS)

Babic, M.; Kljenak, I.

2004-01-01

An experiment on mixing and stratification in the atmosphere of a nuclear power plant containment at accident conditions was simulated with the CFD code CFX4.4. The original experiment was performed in the TOSQAN experimental facility. Simulated nonhomogeneous temperature, species concentration and velocity fields are compared to experimental results. (author)

Reactivation of a tin oxide-containing catalyst

Science.gov (United States)

Upchurch, Billy T. (Inventor); Miller, Irvin M. (Inventor); Brown, Kenneth G. (Inventor); Hess, Robert V. (Inventor); Schryer, David R. (Inventor); Sidney, Barry D. (Inventor); Wood, George M. (Inventor); Paulin, Patricia A. (Inventor)

1989-01-01

A method for the reactivation of a tin oxide-containing catalyst of a CO.sub.2 laser is provided. First, the catalyst is pretreated by a standard procedure. When the catalyst experiences diminished activity during usage, the heated zone surrounding the catalyst is raised to a temperature which is the operating temperature of the laser and 400.degree. C. for approximately one hour. The catalyst is exposed to the same laser gas mixture during this period. The temperature of the heated zone is then lowered to the operating temperature of the CO.sub.2 laser.

Passive cooling containment study

International Nuclear Information System (INIS)

Shin, J.J.; Iotti, R.C.; Wright, R.F.

1993-01-01

Pressure and temperature transients of nuclear reactor containment following postulated loss of coolant accident with a coincident station blackout due to total loss of all alternating current power are studied analytically and experimentally for the full scale NPR (New Production Reactor). All the reactor and containment cooling under this condition would rely on the passive cooling system which removes reactor decay heat and provides emergency core and containment cooling. Containment passive cooling for this study takes place in the annulus between containment steel shell and concrete shield building by natural convection air flow and thermal radiation. Various heat transfer coefficients inside annular air space were investigated by running the modified CONTEMPT code CONTEMPT-NPR. In order to verify proper heat transfer coefficient, temperature, heat flux, and velocity profiles were measured inside annular air space of the test facility which is a 24 foot (7.3m) high, steam heated inner cylinder of three foot (.91m) diameter and five and half foot (1.7m) diameter outer cylinder. Comparison of CONTEMPT-NPR and WGOTHIC was done for reduced scale NPR

A Simple Calorimetric Experiment that Highlights Aspects of Global Heat Retention and Global Warming

Science.gov (United States)

Burley, Joel D.; Johnston, Harold S.

2007-01-01

In this laboratory experiment, general chemistry students measure the heating curves for three different systems: (i) 500 g of room-temperature water heated by a small desk lamp, (ii) 500 g of an ice-water mixture warmed by conduction with room-temperature surroundings, and (iii) 500 g of an ice-water mixture heated by a small desk lamp and by…

Integral Reactor Containment Condensation Model and Experimental Validation

Energy Technology Data Exchange (ETDEWEB)

Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

2016-05-02

This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

Condensation in the presence of noncondensible gases: AP600 containment simulation

Energy Technology Data Exchange (ETDEWEB)

Anderson, M.H.; Corradini, M.L.

1995-09-01

The Westinghouse Electric Corporation has designed an advanced pressurized light water reactor, AP600. This reactor is designed with a passive cooling system to remove sensible and decay heat from the containment. The heat removal path involves condensation heat transfer, aided by natural convective forces generated by buoyancy effects. A one-twelfth scale rectangular slice of the proposed reactor containment was constructed at the University of Wisconsin to simulate conditions anticipated from transients and accidents that may occur in a full scale containment vessel under a variety of conditions. Similitude of the test facility was obtained by considering the appropriate dimensionless group for the natural convective process (modified Froude number) and the aspect ratio (H/R) of the containment vessel. An experimental investigation to determine the heat transfer coefficients associated with condensation on a vertical and horizontal cooled wall (located in the scaled test section) at several different inlet steam flow rates and test section temperatures was conducted. In this series of experiments, the non-condensible mass fraction varied between (0.9-0.4) with corresponding mixture temperatures between 60-90{degrees}C. The heat transfer coefficients of the top horizontal surface varied from (82-296)W/m{sup 2}K and the vertical side heat transfer coefficients varied form (70-269)m{sup 2}K. The results were then compared to boundary layer heat and mass transfer theory by the use of the McAdams correlation for free convection.

Burnout detector design for heat transfer experiments

International Nuclear Information System (INIS)

Dias, H.F.

1992-01-01

This paper describes the design of an burnout detector for heat transfer experiments, applied during tests for optimization of fuel elements for PWR reactors. The burnout detector avoids the fuel rods destruction during the experiments at the Centro de Desenvolvimento da Tecnologia Nuclear. The detector evaluates the temperature changes over the fuel rods in the temperature changes over the fuel rods in the area where the burnout phenomenon could be anticipated. As soon as the phenomenon appears, the system power supply is turned off. The thermal Circuit No. 1, during the experiments, had been composed by nine fuel rods feed parallelly by the same power supply. Fine copper wires had been attached at the centre and at the ends of the fuel rod to take two Wheat stone bridge arms. The detector had been applied across the bridge diagonals, which must be balanced the burnout excursion can be detected as a small but fast increase of the signal over the detector. Large scale experiments had been carried out to compare the resistance bridge performance against a thermocouple attached through the fuel rod wall. These experiments had been showed us the advantages of the first method over the last, because the bridge evaluates the whole fuel rod, while the thermocouple evaluates only the area where it had been attached. (author)

Brine Migration in Heated Salt: Lessons Learned from Field Experiments

Science.gov (United States)

Kuhlman, K. L.; Matteo, E. N.; Mills, M.

2017-12-01

We summarize several interesting brine migration related phenomena hinted at in field experiments from field testing related to salt radioactive waste repositories in Germany and the US. Past heater tests in salt have shown 1) thermal-hydrological-mechanical coupling is quite strong during both heating and cooling; 2) chemical composition of brine evolves during heating, and comprises a mix of several water sources; and 3) acid gas (HCl) generation has been observed during past heater tests and may have multiple mechanisms for formation. We present a heated brine migration test design, formulated with these complexities in mind. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and Engineering Solutions of Sandia LLC, a wholly owned subsidiary of Honeywell International Inc. for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-NA0003525.

Storage chamber for container of radiation-contaminated material

International Nuclear Information System (INIS)

Takakura, Masahide.

1996-01-01

The present invention concerns a storage chamber for containing radiation-contaminated materials in containing tubes and having cooling fluids circulated at the outer side of the containing tubes. The storage chamber comprises a gas supply means connected to the inside of the container tube for supplying a highly heat-conductive gas and a gas exhaustion means for discharging the gas present in the container tube. When containing vessels for radiation-contaminated materials are contained in the container tube, the gases present inside of the container tube is exhausted by means of the gas exhaustion means, and highly heat conductive gases are filled from the gas supply means to the space between the container tube and the containing vessels for the radiation-contaminated materials. When the temperature of the highly heat conductive gas is elevated due to the heat generation of the radiation-contaminated materials, the container tube is heated, and then cooled by the cooling fluid at the outer side of the container tube. In this case, the heat of the radiation-contaminated material-containing vessels is removed by the heat conduction by the highly heat conductive gas to reduce temperature gradient between the containing vessels and the containing tube. This can enhance the cooling effect. (T.M.)

Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

International Nuclear Information System (INIS)

Spring, J.P.; McLaughlin, D.M.

2006-01-01

Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7 x 7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 deg. C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local

Containing method for spent fuel and spent fuel containing vessel

International Nuclear Information System (INIS)

Maekawa, Hiromichi; Hanada, Yoshine.

1996-01-01

Upon containing spent fuels, a metal vessel main body and a support spacer having fuel containing holes are provided. The support spacer is disposed in the inside of the metal vessel main body, and spent fuel assemblies are loaded in the fuel containing holes. Then, a lid is welded at the opening of the metal vessel main body to provide a sealing state. In this state, heat released from the spent fuel assemblies is transferred to the wall of the metal vessel main body via the support spacer. Since the support spacer has a greater heat conductivity than gases, heat of the spent fuel assemblies tends to be released to the outside, thereby capable of removing heat of the spent fuel assemblies effectively. In addition, since the surfaces of the spent fuel assemblies are in contact with the inner surface of the fuel containing holes of the support spacer, impact-resistance and earthquake-resistance are ensured, and radiation from the spent fuel assemblies is decayed by passing through the layer of the support spacer. (T.M.)

Coupling between particle and heat transport during power modulation experiments in Tore Supra

International Nuclear Information System (INIS)

Zou, X.L.; Giruzzi, G.; Artaud, J.F.; Bouquey, F.; Bremond, S.; Clary, J.; Darbos, C.; Eury, S.P.; Lennholm, M.; Magne, R.; Segui, J.L.

2004-01-01

Power modulations are a powerful tool often used to investigate heat transport processes in tokamaks. In some situations, this could also be an interesting method for the investigation of the particle transport due to the anomalous pinch. Low frequency (∼ 1 Hz) power modulation experiments, using both electron cyclotron resonance heating (ECRH) and ion cyclotron resonance heating (ICRH), have been performed in the Tore Supra tokamak. Strong coupling has been observed between the temperature and density modulations during the low frequency ECRH and ICRH modulation experiments. It has been shown that mechanisms as outgassing, Ware pinch effect, curvature driven pinch are not likely to be responsible for this density modulation. Because of its dependence on temperature or temperature gradient, the thermodiffusion is a serious candidate to be the driving source for this density modulation. This analysis shows that low frequency power modulation experiments have a great potential for the investigation of the anomalous particle pinch in tokamaks. Future plans will include the use of more precise density profile measurements using X-mode reflectometry

Coupling between particle and heat transport during power modulation experiments in Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Zou, X.L.; Giruzzi, G.; Artaud, J.F.; Bouquey, F.; Bremond, S.; Clary, J.; Darbos, C.; Eury, S.P.; Lennholm, M.; Magne, R.; Segui, J.L

2004-07-01

Power modulations are a powerful tool often used to investigate heat transport processes in tokamaks. In some situations, this could also be an interesting method for the investigation of the particle transport due to the anomalous pinch. Low frequency ({approx} 1 Hz) power modulation experiments, using both electron cyclotron resonance heating (ECRH) and ion cyclotron resonance heating (ICRH), have been performed in the Tore Supra tokamak. Strong coupling has been observed between the temperature and density modulations during the low frequency ECRH and ICRH modulation experiments. It has been shown that mechanisms as outgassing, Ware pinch effect, curvature driven pinch are not likely to be responsible for this density modulation. Because of its dependence on temperature or temperature gradient, the thermodiffusion is a serious candidate to be the driving source for this density modulation. This analysis shows that low frequency power modulation experiments have a great potential for the investigation of the anomalous particle pinch in tokamaks. Future plans will include the use of more precise density profile measurements using X-mode reflectometry.

High Harmonic Fast Wave Heating Experiments on NSTX

International Nuclear Information System (INIS)

Wilson, J.R.; Bell, R.; Bitter, M.; Bonoli, P.

2000-01-01

A radio frequency (rf) system has been installed on the National Spherical Torus Experiment (NSTX) with the aim of heating the plasma and driving plasma current. The system consists of six rf transmitters, a twelve element antenna and associated transmission line components to distribute and couple the power from the transmitters to the antenna elements in a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, power levels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum of the rf waves has been selected to heat electrons via Landau damping and transit time magnetic pumping. The electron temperature has been observed to increase from 400 to 900 eV with little change in plasma density resulting in a plasma stored energy of 59 kJ and a toroidal beta, bT , =10% and bn = 2.7

High harmonic fast wave heating experiments on NSTX

International Nuclear Information System (INIS)

Wilson, J.R.; Bell, R.; Bitter, M.

2001-01-01

A radio frequency (rf) system has been installed on the National Spherical Torus Experiment (NSTX) with the aim of heating the plasma and driving plasma current. The system consists of six rf transmitters, a twelve element antenna and associated transmission line components to distribute and couple the power from the transmitters to the antenna elements in a fashion to allow control of the antenna toroidal wavenumber spectrum. To date, power levels up to 3.85 MW have been applied to the NSTX plasmas. The frequency and spectrum of the rf waves has been selected to heat electrons via Landau damping and transit time magnetic pumping. The electron temperature has been observed to increase from 400 to 900 eV with little change in plasma density resulting in a plasma stored energy of 59 kJ , a toroidal beta, β T =10% and a normalized beta, β n =2.7. (author)

Effect of prolonged heating on the asphalt-aggregate bond strength of HMA containing liquid anti-strip additives

Science.gov (United States)

2008-09-01

In this study, an attempt was made to determine the effect of prolonged heating on the bond strength between : aggregate and asphalt that contained anti-strip additives (LOF 6500 and Morelife 2200). On account of the : substantial decrease of anti-st...

Experiments on nucleate boiling heat transfer with a highly-wetting dielectric fluid

International Nuclear Information System (INIS)

You, S.M.; Simon, T.W.; Bar-Cohen, A.

1990-01-01

This paper reports on experiments on pool boiling heat transfer in an electronic cooling fluid (Fluorinert, FC-72) that were conducted using a 0.51 mm diameter cylindrical heater. The effects of pressure, subcooling and dissolved gas content on nucleate boiling heat transfer are investigated. When boiling with dissolved gas in the bulk fluid, the fluid in the vicinity of the heating element appears to be liberated of dissolved gas by boiling. Thus, boiling under these conditions appears to be similar to subcooled boiling without dissolved gas. Nucleate boiling hysteresis is observed for subcooled and gassy-subcooled situations

Technical specification improvements to containment heat removal and emergency core cooling systems: Final report

International Nuclear Information System (INIS)

Sullivan, W.P.; Ha, C.; Pentzien, D.C.; Visweswaran, S.

1988-07-01

This report presents the results of an analysis for technical specification improvements to the emergency core cooling systems (ECCS) and containment heat removal systems (EPRI Research Project 2142-3). The objective of this project is to further develop a reliability- and risk-based methodology to provide improvements by considering groups of surveillance test intervals and allowed out-of-service times jointly. This was done for the technical specifications for the ECCS, containment heat removal equipment, and supporting systems of a boiling water reactor plant. The project (1) developed a methodology for optimizing groups of surveillance test intervals and allowed out-of-service times jointly, (2) applied the methodology in a case study of a specific operating plant, Hatch-2, and (3) evaluated benefits of the application. The results of the case study demonstrate that beneficial technical specification improvements can be realized with application of the methodology. By tightening a small group of sensitive surveillance test intervals (STIs) and allowed out-of-service times (AOTs), a larger group of less sensitive STIs and AOTs can be extended resulting in an overall plant operating cost improvement without reducing the plant safety. The reliability- and risk-based methodology and results from this project can be effectively applied for technical specification improvements at other operating plants

Passive containment system

International Nuclear Information System (INIS)

Kleimola, F.W.

1977-01-01

Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

Atlas of the global distribution of atmospheric heating during the global weather experiment

Science.gov (United States)

Schaack, Todd K.; Johnson, Donald R.

1991-01-01

Global distributions of atmospheric heating for the annual cycle of the Global Weather Experiment are estimated from the European Centre for Medium-Range Weather Forecasts (ECMWF) Level 3b data set. Distributions of monthly, seasonally, and annually averaged heating are presented for isentropic and isobaric layers within the troposphere and for the troposphere as a whole. The distributions depict a large-scale structure of atmospheric heating that appears spatially and temporally consistent with known features of the global circulation and the seasonal evolution.

Heat wave experiments on the W7-AS stellarator

International Nuclear Information System (INIS)

Hartfuss, H.J.; Erckmann, V.; Gasparino, U.; Giannone, L.; Maassberg, H.; Tutter, M.

1993-01-01

Power modulation with well localized ECRH power deposition at both 70 and 140 GHz, has been used to generate temperature perturbations which propagate away from the deposition region. Radiometry of the ECE is used to diagnose the generated temperature perturbation as a function of distance to the deposition zone. The decay of the amplitude and the delay of the wave provide the information to determine the electron thermal diffusivity. This value is then compared with the one derived from a global power balance. It is found that both values agree with the error bars. The technique has also been applied in recent experiments during L-H-mode transitions in W7-AS demonstrating a significant reduction in the effective heat diffusivity in the plasma core during the H-phase. The modulated ECRH causes a modulation of the Shafranov shift. Interference of the prompt shift with the heat wave results in an apparent asymmetry of the decay length of the heat wave with respect to the plasma centre. (orig.)

Heat transfer from rough surfaces

International Nuclear Information System (INIS)

Dalle Donne, M.

1977-01-01

Artificial roughness is often used in nuclear reactors to improve the thermal performance of the fuel elements. Although these are made up of clusters of rods, the experiments to measure the heat transfer and friction coefficients of roughness are performed with single rods contained in smooth tubes. This work illustrated a new transformation method to obtain data applicable to reactor fuel elements from these annulus experiments. New experimental friction data are presented for ten rods, each with a different artificial roughness made up of two-dimensional rectangular ribs. For each rod four tests have been performed, each in a different outer smooth tube. For two of these rods, each for two different outer tubes, heat transfer data are also given. The friction and heat transfer data, transformed with the present method, are correlated by simple equations. In the paper, these equations are applied to a case typical for a Gas Cooled Fast Reactor fuel element. (orig.) [de

Research Proposal for the Design and Engineering Phase of a Solar Heating and Cooling System Experiment at the Warner Robins Public Library, Warner Robins, Georgia. Submitted to the United States Energy Research and Development Administration.

Science.gov (United States)

Phillips, Warren H.; And Others

A number of reasons are advanced to include a solar heating and cooling experiment in a library building. The unique aspects of the experiment are to be a seasonally adjustable collector tilt and testing of a new generation of absorption air conditioners. After a brief description of the proposed experiment, the proposal contains forms filed by…

Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

International Nuclear Information System (INIS)

Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

1999-01-01

A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

Review on experiments relating to primary containment vessel failure

International Nuclear Information System (INIS)

Suzuki, Hiroyuki; Okada, Hidetoshi; Uchida, Sunsuke; Naitoh, Masanori

2015-01-01

Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries. (author)

Application of a two-cell adiabatic model for direct containment heating to the ABB C-E system 80+ ALWR

International Nuclear Information System (INIS)

Schneider, R.E.; Sherry, R.R.

1993-01-01

During certain severe reactor accidents, such as those initiated by a station blackout or small-break loss of coolant accident (LOCA) degradation of the reactor core can take place while the reactor coolant system remains pressurized. If unmitigated, core materials will melt and relocate to the lower regions of the reactor pressure vessel and ultimately melt through the reactor pressure vessel (RPV) lower head. Once the RPV is breached, core debris will be ejected from the RPV and entrained from the reactor cavity by the high velocity gases blowing down from the reactor vessel. During the entrainment process, metallic constituents of the ejected material, principally zirconium and steel, exothermically react with oxygen and steam to generate chemical energy and (in the case of reactions with steam) hydrogen. Concomitant with the high pressure melt ejection (HPME) process, there is the potential for hydrogen combustion and vaporization of available water. The sensible heat loss to the containment atmosphere and the associated processes are typically referred to as direct containment heating (DCH). If large quantities of energy from the corium and corium-steam reactions are transferred directly to the containment atmosphere, the containment may pressurize to a point where failure is possible. Since the containment threat is coincident with vessel breach, relatively high containment radiation releases would be expected from this type of containment failure

CEC thermal-hydraulic benchmark exercise on Fiploc verification experiment F2 in Battelle model containment. Experimental phases 2, 3 and 4. Results of comparisons

International Nuclear Information System (INIS)

Fischer, K.; Schall, M.; Wolf, L.

1993-01-01

The present final report comprises the major results of Phase II of the CEC thermal-hydraulic benchmark exercise on Fiploc verification experiment F2 in the Battelle model containment, experimental phases 2, 3 and 4, which was organized and sponsored by the Commission of the European Communities for the purpose of furthering the understanding and analysis of long-term thermal-hydraulic phenomena inside containments during and after severe core accidents. This benchmark exercise received high European attention with eight organizations from six countries participating with eight computer codes during phase 2. Altogether 18 results from computer code runs were supplied by the participants and constitute the basis for comparisons with the experimental data contained in this publication. This reflects both the high technical interest in, as well as the complexity of, this CEC exercise. Major comparison results between computations and data are reported on all important quantities relevant for containment analyses during long-term transients. These comparisons comprise pressure, steam and air content, velocities and their directions, heat transfer coefficients and saturation ratios. Agreements and disagreements are discussed for each participating code/institution, conclusions drawn and recommendations provided. The phase 2 CEC benchmark exercise provided an up-to-date state-of-the-art status review of the thermal-hydraulic capabilities of present computer codes for containment analyses. This exercise has shown that all of the participating codes can simulate the important global features of the experiment correctly, like: temperature stratification, pressure and leakage, heat transfer to structures, relative humidity, collection of sump water. Several weaknesses of individual codes were identified, and this may help to promote their development. As a general conclusion it may be said that while there is still a wide area of necessary extensions and improvements, the

Design and experiment of a new solar air heating collector

International Nuclear Information System (INIS)

Shams, S.M.N.; Mc Keever, M.; Mc Cormack, S.; Norton, B.

2016-01-01

This paper presents the design and experiment of a CTAH (Concentrating Transpired Air Heating) system. A newly designed solar air heating collector comprised of an inverted perforated absorber and an asymmetric compound parabolic concentrator was applied to increase the intensity of solar radiation incident on the perforated absorber. An extensive literature review was carried out to find the vital factors to improve optical and thermal efficiency of solar air heating systems. A stationary optical concentrator has been designed and experimented. Experimental thermal efficiency remained high at higher air flow rates. The average thermal efficiency was found to be approximately 55%–65% with average radiation above 400 W/m"2 for flow rates in the range of 0.03 kg/s/m"2 to 0.09 kg/s/m"2. Experimental results at air flow rates of 0.03 kg/s/m"2 and 0.09 kg/s/m"2 showed temperature rise of 38 °C and 19.6 °C respectively at a solar radiation intensity of 1000 W/m"2. A comparative performance study shows the thermal performance of CTAH. As the absorber of the CTAH facing downward, it avoids radiation loss and the perforated absorber with tertiary concentrator reduces thermal losses from the system. - Highlights: • Literature review was carried out to improve SAH system performance. • Optimisation factors were optical efficiency; heat loss, weight and cost. • Concentrator was designed to concentrate radiation for 6–7 h. • The highest efficiency of CTAH can be 73%. • It can work as efficient as 60% for a temperature rise of 70 °C.

Implementation of Exhaust Gas Recirculation for Double Stage Waste Heat Recovery System on Large Container Vessel

DEFF Research Database (Denmark)

Andreasen, Morten; Marissal, Matthieu; Sørensen, Kim

2014-01-01

Concerned to push ships to have a lower impact on the environment, the International Maritime Organization are implementing stricter regulation of NOx and SOx emissions, called Tier III, within emission control areas (ECAs). Waste Heat Recovery Systems (WHRS) on container ships consist...... of recovering some of the waste heat from the exhaust gas. This heat is converted into electrical energy used on-board instead of using auxiliary engines. Exhaust Gas Recirculation (EGR) systems, are recirculating a part of the exhaust gas through the engine combustion chamber to reduce emissions. WHRS combined...... with EGR is a potential way to improve system efficiency while reducing emissions. This paper investigates the feasibility of combining the two systems. EGR dilutes the fuel, lowering the combustion temperature and thereby the formation of NOx, to reach Tier III limitation. A double stage WHRS is set up...

Heat Shock Protein-Inducing Property of Diarylheptanoid Containing Chalcone Moiety from Alpinia katsumadai

Directory of Open Access Journals (Sweden)

Joo-Won Nam

2017-10-01

Full Text Available A new diarylheptanoid containing a chalcone moiety, katsumain H (1, was isolated from the seeds of Alpinia katsumadai. The structure was elucidated using a combination of 1D/2D NMR spectroscopy and mass spectrometry data analysis. The absolute configurations of C-3, C-5, and C-7 in 1 were assigned based on its optical rotation and after comparing its NMR chemical shifts with those of its diastereoisomers, katsumain E and katsumain F, which were previously isolated from this plant and characterized. In this study, the stimulatory effects of compounds 1 and 2 were evaluated on heat shock factor 1 (HSF1, heat shock protein 27 (HSP27, and HSP70. Compounds 1 and 2 increased the expression of HSF1 (1.056- and 1.200-fold, respectively, HSP27 (1.312- and 1.242-fold, respectively, and HSP70 (1.234- and 1.271-fold, respectively, without increased cytotoxicity.

Exploration of high harmonic fast wave heating on the National Spherical Torus Experiment

International Nuclear Information System (INIS)

Wilson, J.R.; Bell, R.E.; Bernabei, S.; Bitter, M.; Gates, D.; Hosea, J.; Le Blanc, B.; Medley, S.; Menard, J.; Mueller, D.; Ono, M.; Phillips, C.K.; Rosenberg, A.; Bonoli, P.; Mau, T.K.; Pinsker, R.I.; Raman, R.; Ryan, P.; Swain, D.; Wilgen, J.

2003-01-01

High harmonic fast wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [M. Ono, S. M. Kaye, S. Neumeyer et al., in Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999 (IEEE, Piscataway, NJ, 1999), p. 53] is such a device. An rf heating system has been installed on the NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to explore the energy transport properties of NSTX plasmas, to create H-mode discharges with a large fraction of bootstrap current and to control the plasma current profile during the early stages of the discharge

Exploration of High Harmonic Fast Wave Heating on the National Spherical Torus Experiment

International Nuclear Information System (INIS)

Wilson, J.R.; Bell, R.E.; Bernabei, S.; Bitter, M.; Bonoli, P.; Gates, D.; Hosea, J.; LeBlanc, B.; Mau, T.K.; Medley, S.; Menard, J.; Mueller, D.; Ono, M.; Phillips, C.K.; Pinsker, R.I.; Raman, R.; Rosenberg, A.; Ryan, P.; Sabbagh, S.; Stutman, D.; Swain, D.; Takase, Y.; Wilgen, J.

2003-01-01

High Harmonic Fast Wave (HHFW) heating has been proposed as a particularly attractive means for plasma heating and current drive in the high-beta plasmas that are achievable in spherical torus (ST) devices. The National Spherical Torus Experiment (NSTX) [Ono, M., Kaye, S.M., Neumeyer, S., et al., Proceedings, 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque, 1999, (IEEE, Piscataway, NJ (1999), p. 53.)] is such a device. An radio-frequency (rf) heating system has been installed on NSTX to explore the physics of HHFW heating, current drive via rf waves and for use as a tool to demonstrate the attractiveness of the ST concept as a fusion device. To date, experiments have demonstrated many of the theoretical predictions for HHFW. In particular, strong wave absorption on electrons over a wide range of plasma parameters and wave parallel phase velocities, wave acceleration of energetic ions, and indications of current drive for directed wave spectra have been observed. In addition HHFW heating has been used to explore the energy transport properties of NSTX plasmas, to create H-mode (high-confinement mode) discharges with a large fraction of bootstrap current and to control the plasma current profile during the early stages of the discharge

JET ({sup 3}He)-D scenarios relying on RF heating: survey of selected recent experiments

Energy Technology Data Exchange (ETDEWEB)

Van Eester, D; Lerche, E; Andrew, Y; Biewer, T M; Casati, A; Crombe, K; De la Luna, E; Ericsson, G; Felton, R; Giacomelli, L; Giroud, C; Hawkes, N; Hellesen, C; Hjalmarsson, A; Joffrin, E; Kaellne, J; Kiptily, V; Lomas, P; Mantica, P; Marinoni, A [JET-EFDA Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)] (and others)

2009-04-15

Recent JET experiments have been devoted to the study of ({sup 3}He)-D plasmas involving radio frequency (RF) heating. This paper starts by discussing the RF heating efficiency theoretically expected in such plasmas, covering both relevant aspects of wave and of particle dynamics. Then it gives a concise summary of the main conclusions drawn from recent experiments that were either focusing on studying RF heating physics aspects or that were adopting RF heating as a tool to study plasma behavior. Depending on the minority concentration chosen, different physical phenomena are observed. At very low concentration (X[{sup 3}He] < 1%), energetic tails are formed which trigger MHD activity and result in loss of fast particles. Alfven cascades were observed and gamma ray tomography indirectly shows the impact of sawtooth crashes on the fast particle orbits. Low concentration (X[{sup 3}He] < 10%) favors minority heating while for X[{sup 3}He] >> 10% electron mode conversion damping becomes dominant. Evidence for the Fuchs et al standing wave effect (Fuchs et al 1995 Phys. Plasmas 2 1637-47) on the absorption is presented. RF induced deuterium tails were observed in mode conversion experiments with large X[{sup 3}He] ({approx}18%). As tentative modeling shows, the formation of these tails can be explained as a consequence of wave power absorption by neutral beam particles that efficiently interact with the waves well away from the cold D cyclotron resonance position as a result of their substantial Doppler shift. As both ion and electron RF power deposition profiles in ({sup 3}He)-D plasmas are fairly narrow-giving rise to localized heat sources-the RF heating method is an ideal tool for performing transport studies. Various of the experiments discussed here were done in plasmas with internal transport barriers (ITBs). ITBs are identified as regions with locally reduced diffusivity, where poloidal spinning up of the plasma is observed. The present know-how on the role of

ANL ITER high-heat-flux blanket-module heat transfer experiments

International Nuclear Information System (INIS)

Kasza, K.E.

1992-02-01

An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER

Heat transfer characteristics of building walls using phase change material

Science.gov (United States)

Irsyad, M.; Pasek, A. D.; Indartono, Y. S.; Pratomo, A. W.

2017-03-01

Minimizing energy consumption in air conditioning system can be done with reducing the cooling load in a room. Heat from solar radiation which passes through the wall increases the cooling load. Utilization of phase change material on walls is expected to decrease the heat rate by storing energy when the phase change process takes place. The stored energy is released when the ambient temperature is low. Temperature differences at noon and evening can be utilized as discharging and charging cycles. This study examines the characteristics of heat transfer in walls using phase change material (PCM) in the form of encapsulation and using the sleeve as well. Heat transfer of bricks containing encapsulated PCM, tested the storage and released the heat on the walls of the building models were evaluated in this study. Experiments of heat transfer on brick consist of time that is needed for heat transfer and thermal conductivity test as well. Experiments were conducted on a wall coated by PCM which was exposed on a day and night cycle to analyze the heat storage and heat release. PCM used in these experiments was coconut oil. The measured parameter is the temperature at some points in the brick, walls and ambient temperature as well. The results showed that the use of encapsulation on an empty brick can increase the time for thermal heat transfer. Thermal conductivity values of a brick containing encapsulated PCM was lower than hollow bricks, where each value was 1.3 W/m.K and 1.6 W/m.K. While the process of heat absorption takes place from 7:00 am to 06:00 pm, and the release of heat runs from 10:00 pm to 7:00 am. The use of this PCM layer can reduce the surface temperature of the walls of an average of 2°C and slows the heat into the room.

HTCC - a heat transfer model for gas-steam mixtures

International Nuclear Information System (INIS)

Papadimitriou, P.

1983-01-01

The mathematical model HTCC (Heat Transfer Coefficient in Containment) has been developed for RALOC after a loss-of-coolant accident in order to determine the local heat transfer coefficients for transfer between the containment atmosphere and the walls of the reactor building. The model considers the current values of room and wall temperature, the concentration of steam and non-condensible gases, geometry data and those of fluid dynamics together with thermodynamic parameters and from these determines the heat transfer mechanisms due to convection, radiation and condensation. The HTCC is implemented in the RALOC program. Comparative analyses of computed temperature profiles, for HEDL Standard problems A and B on hydrogen distribution, and of computed temperature profiles determined during the heat-up phase in the CSE-A5 experiment show a good agreement with experimental data. (orig.) [de

AREVA’s Containment Venting Technologies and Experience Worldwide

Energy Technology Data Exchange (ETDEWEB)

Welker, M.

2015-07-01

The AREVA Filtered Containment Venting System (FCVS) is a product family that minimizes the environmental impact in case of a severe accident in a nuclear power plant (NPP). Our experience is based on a large-scale test and qualification program as well as on the design, licensing and installation of more than 80 projects worldwide. The product family provides flexibility regarding the adaptation to respective accident scenarios, applicable codes and standards, seismic design, supply chain, implementation and localization. AREVA has broad experience of managing fleet supplies, successful support of licensing and cooperating with original equipment manufacturers (OEMs) of pressurized and boiling water reactors (PWR and BWR). (Author)

A phenomenological explanation for the anomalous ion heating observed in the JET alpha-heating experiment of 1997

Science.gov (United States)

Testa, D.; Albergante, M.

2012-08-01

In the so-called ‘alpha-heating’ experiment performed on the JET tokamak during the deuterium-tritium campaign of 1997, the ion temperature was found to be far exceeding (both in absolute value and in its rise time) the level that could have been expected from direct collisional heating by the fusion-born alpha particles themselves and energy equipartition with the electrons. To date, no explanation has been put forward for this long standing puzzle, despite much work having been performed on this subject in the early 2000s. Two analysis methods that have recently become available have been employed to re-analyse these observations of an anomalous ion heating. First, an algorithm based on the sparse representation of signals has been used to analyse magnetic, reflectometry and electron-cyclotron emission measurements of the turbulence spectra in the drift-wave range of frequencies. This analysis has then been complemented with turbulence simulations performed with the GENE code. We find, both experimentally and in the simulations, that the presence of a minority, but sufficiently large, population of fusion-born alpha particles that have not yet fully thermalized stabilizes the turbulence in the ion-drift direction, but practically does not affect the turbulence in the electron-drift direction. We link such stabilization of the ion-drift-wave turbulence to the increase in the ion temperature above the level achieved in similar discharges that did not have (at all or enough) alpha particles. When the fusion-born alpha particles have fully thermalized, the turbulence spectrum in the ion-drift direction reappears at somewhat larger amplitudes, which we link to the ensuing reduction in the ion temperature. This phenomenological dynamics fully corresponds to the actual experimental observations. By taking into account an effect of the alpha particles that had not been previously considered, our new analysis finally presents a phenomenological explanation for the so

A phenomenological explanation for the anomalous ion heating observed in the JET alpha-heating experiment of 1997

International Nuclear Information System (INIS)

Testa, D.; Albergante, M.

2012-01-01

In the so-called ‘alpha-heating’ experiment performed on the JET tokamak during the deuterium–tritium campaign of 1997, the ion temperature was found to be far exceeding (both in absolute value and in its rise time) the level that could have been expected from direct collisional heating by the fusion-born alpha particles themselves and energy equipartition with the electrons. To date, no explanation has been put forward for this long standing puzzle, despite much work having been performed on this subject in the early 2000s. Two analysis methods that have recently become available have been employed to re-analyse these observations of an anomalous ion heating. First, an algorithm based on the sparse representation of signals has been used to analyse magnetic, reflectometry and electron-cyclotron emission measurements of the turbulence spectra in the drift-wave range of frequencies. This analysis has then been complemented with turbulence simulations performed with the GENE code. We find, both experimentally and in the simulations, that the presence of a minority, but sufficiently large, population of fusion-born alpha particles that have not yet fully thermalized stabilizes the turbulence in the ion-drift direction, but practically does not affect the turbulence in the electron-drift direction. We link such stabilization of the ion-drift-wave turbulence to the increase in the ion temperature above the level achieved in similar discharges that did not have (at all or enough) alpha particles. When the fusion-born alpha particles have fully thermalized, the turbulence spectrum in the ion-drift direction reappears at somewhat larger amplitudes, which we link to the ensuing reduction in the ion temperature. This phenomenological dynamics fully corresponds to the actual experimental observations. By taking into account an effect of the alpha particles that had not been previously considered, our new analysis finally presents a phenomenological explanation for the

Heat transfer characteristics of UF6 in a container heated from outer surface. Pt. 1. Thermal hydraulic analysis method taking account of phase change and volume expansion

International Nuclear Information System (INIS)

Wataru, Masumi; Gomi, Yoshio; Yamakawa, Hidetsugu; Tsumune, Daisuke

1995-01-01

Natural UF6 is transported in a steel container from foreign countries to the enrichment plant in Japan. If the container meets fire accident, it is heated by fire (800degC) and rupture of the container may occur. For the safety point of view, it is necessary to know whether rupture occurs or not. Because UF6 has a radiological and chemical hazards, it is difficult to perform a demonstration test with UF6. So thermal calculation method has to be developed. The rupture is caused by UF6 gaseous pressure or volume expansion of liquid UF6. To know time history of internal pressure and temperature distribution in the container, it is important to evaluate thermal phenomena of UF6. When UF6 is heated, it changes from solid to liquid or gas at low temperature (64degC) and then its volume expands little by little. In this study, thermal calculation method has been developed taking phase change and thermal expansion of UF6 into account. In the calculation, a two-dimensional model is adopted and natural convection of liquid UF6 is analyzed. As a result of this study, numerical solutions have been obtained taking phase change and volume expansion into account. (author)

Deposition of aerosols formed by HCDA due to decay heat transport in inner containment atmospheres

International Nuclear Information System (INIS)

Vate, J.F. van de

1976-01-01

Coupling of decay heat transfer by aerosol-laden inner containment atmospheres with aerosol deposition from such atmospheres leads to useful and simple models for calculation of the time dependence of the aerosol mass concentration. Special attention is given to thermophoretic deposition (dry case) and condensation followed by gravitational deposition (wet case). Attractive features of the models are: 1) coagulation can be omitted and therefore complicated and doubtful calculations on coagulation are avoided, 2) material and particle size of the aerosol are not important for the aerosol decay rate, 3) the aerosol decay rate is related to the decay heat production which is known function of time, and the relevant part of it must be assessed usually for other purposes as well. (orig.) [de

Four decades of working experience of Cirus primary cooling water heat exchangers

International Nuclear Information System (INIS)

Dubey, P.K.; Ullas, O.P.; Rao, D.V.H.; Zope, A.K.; Kharpate, A.V.

2006-01-01

CIRUS is a 40 MW (Th.) research reactor, commissioned in the year 1960. The reactor has natural uranium fuel rods, heavy water as moderator, demineralised water (DM water) as primary coolant, and seawater as secondary coolant. There are six Heat Exchangers in the primary cooling water (PCW) system. Five of them are required for the normal operation of the reactor and one is kept stand by. DM water flows on the shell side of the heat exchanger in two passes. Seawater is used as coolant on the tube side of the heat exchangers in four passes. Cirus has been in operation for around 41 years excluding refurbishment period. During these four decades of reactor operation, PCW heat exchangers have experienced many failures and undergone many modifications in the circuit for ensuring better performance. This paper tries to capture the essence of working experiences with PCW heat exchangers, various problems faced, remedial measures taken during those four decades of reactor operation. (author)

Design of an experiment to measure the decay heat of an irradiated PWR fuel: MERCI experiment; Conception d'une experience de mesure de la puissance residuelle d'un combustible irradie: l'experience MERCI

Energy Technology Data Exchange (ETDEWEB)

Bourganel, St

2002-11-01

After a reactor shutdown, a significant quantity of energy known as 'decay heat' continues to be generated from the irradiated fuel. This heat source is due to the disintegration energy of fission products and actinides. Decay heat determination of an irradiated fuel is of the utmost importance for safety analysis as the design cooling systems, spent fuel transport, or handling. Furthermore, the uncertainty on decay heat has a straight economic impact. The unloading fuel spent time is an example. The purpose of MERCI experiment (irradiated fuel decay heat measurement) consists in qualifying computer codes, particularly the DARWIN code system developed by the CEA in relation to industrial organizations, as EDF, FRAMATOME and COGEMA. To achieve this goal, a UOX fuel is irradiated in the vicinity of the OSIRIS research reactor, and then the decay heat is measured by using a calorimeter. The objective is to reduce the decay heat uncertainties from 8% to 3 or 4% at short cooling times. A full simulation on computer of the MERCI experiment has been achieved: fuel irradiation analysis is performed using transport code TRIPOLI4 and evolution code DARWIN/PEPIN2, and heat transfer with CASTEM2000 code. The results obtained are used for the design of this experiment. Moreover, we propose a calibration procedure decreasing the influence of uncertainty measurements and an interpretation method of the experimental results and evaluation of associated uncertainties. (author)

Heat Transfer by Thermo-capillary Convection -Sounding Rocket COMPERE Experiment SOURCE

Science.gov (United States)

Dreyer, Michael; Fuhrmann, Eckart

The sounding rocket COMPERE experiment SOURCE was successfully flown on MASER 11, launched in Kiruna (ESRANGE), May 15th, 2008. SOURCE has been intended to partly ful-fill the scientific objectives of the European Space Agency (ESA) Microgravity Applications Program (MAP) project AO-2004-111 (Convective boiling and condensation). Three parties of principle investigators have been involved to design the experiment set-up: ZARM for thermo-capillary flows, IMFT (Toulouse, France) for boiling studies, EADS Astrium (Bremen, Ger-many) for depressurization. The topic of this paper is to study the effect of wall heat flux on the contact line of the free liquid surface and to obtain a correlation for a convective heat trans-fer coefficient. The experiment has been conducted along a predefined time line. A preheating sequence at ground was the first operation to achieve a well defined temperature evolution within the test cell and its environment inside the rocket. Nearly one minute after launch, the pressurized test cell was filled with the test liquid HFE-7000 until a certain fill level was reached. Then the free surface could be observed for 120 s without distortion. Afterwards, the first depressurization was started to induce subcooled boiling, the second one to start saturated boiling. The data from the flight consists of video images and temperature measurements in the liquid, the solid, and the gaseous phase. Data analysis provides the surface shape versus time and the corresponding apparent contact angle. Computational analysis provides information for the determination of the heat transfer coefficient in a compensated gravity environment where a flow is caused by the temperature difference between the hot wall and the cold liquid. The paper will deliver correlations for the effective contact angle and the heat transfer coefficient as a function of the relevant dimensionsless parameters as well as physical explanations for the observed behavior. The data will be used

Heat transfer on liquid-liquid interface of molten-metal and water

International Nuclear Information System (INIS)

Tanaka, T.; Saito, Yasushi; Mishima, Kaichiro

2001-01-01

Molten-core pool had been formed in the lower-head of TMI-2 pressure vessel at the severe accident. The lower head, however, didn't receive any damage by reactor core cooling. Heat transfer at outside of the lower head and boiling heat transfer at liquid-liquid interface of molten-metal and water, however, are important for initial cooling process of the molten-core pool. The heat transfer experiments for the liquid-liquid interface of molten-metal and water are carried out over the range of natural convection to film boiling region. Phenomenon on the heat transfer experiments are visualized by using of high speed video camera. Wood's metal and U-alloy 78 are used as molten-metal. The test section of the experiments consists of a copper block with heater, wood's metal, and water. Three thermocouple probes are used for temperature measurement of water side and the molten-metal side. Stability of the liquid-liquid interface is depended on the wetness of container wall for molten metal and the temperature distribution of the interface. Entrainment phenomena of molten-metal occurs by a fluctuation of the interface after boiling on the container wall surface. The boiling curves obtained from the liquid-liquid interface experiments are agree with the nucleate boiling and the film boiling correlations of solid-liquid system. (Suetake, M.)

Heat pipe

International Nuclear Information System (INIS)

Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.

1986-01-01

A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)

ICRF heating experiments in JFT-2 tokamak

International Nuclear Information System (INIS)

Matsumoto, Hiroshi

1986-01-01

This is an experimental study of ICRF heating on JFT-2 Tokamak in Japan Atomic Energy Research Institute. In this study, we first clarified physical and engineering problems of ICRF heating of tokamak plasma. Next, we optimized the design of the ICRF heating system, and the plasma parameters for the heating. Finally, we could demonstrate a high efficiency of this additional heating method by launching RF power which is two or three times as large as an ohmic input power to a plasma. And we achieved following things. (1) We optimized a design of an antenna, and we improved a durability of the system for high voltage. With the result that we achieved the maximum power density on an antenna. (2) We demonstrated that electron heating regime and ion heating regime can be easily accessed by controlling plasma parameters. Also we found the optimum heating conditions in each heating regime. (3) We experimentally clarified the production mechanism of impurities during ICRF heating. We could reduce the influx of metal impurity ions to a plasma by employing low z materials for limiters and antenna shields. Consequently, we improved a heating efficiency of electrons. Next, we studied a power balance of plasma during ICRF heating, and we could compare heating characteristics of ICRF with other additional heatings on JFT-2. (author)

A Fresnel collector process heat experiment at Capitol Concrete Products

Science.gov (United States)

Hauger, J. S.

1981-01-01

An experiment is planned, conducted and evaluated to determine the feasibility of using a Power Kinetics' Fresnel concentrator to provide process heat in an industrial environment. The plant provides process steam at 50 to 60 psig to two autoclaves for curing masonry blocks. When steam is not required, the plant preheats hot water for later use. A second system is installed at the Jet Propulsion Laboratory parabolic dish test site for hardware validation and experiment control. Experiment design allows for the extrapolation of results to varying demands for steam and hot water, and includes a consideration of some socio-technical factors such as the impact on production scheduling of diurnal variations in energy availability.

Heat Transfer Reactor Experiment (HTRE)-3 Container Storage Unit Resource Conservation Recovery Act closure plan

International Nuclear Information System (INIS)

Spry, M.J.

1992-11-01

This document describes the closure of the HTRE-3 Container Storage Unit under the requirements of the Resource Conservation and Recovery Act. The unit's location, size, history, and current status are described. The document also summarizes the decontamination and decommissioning efforts performed in 1983 and provides an estimate of,waste residues remaining in the HTRE-3 assembly. A risk evaluation was performed that demonstrates that the residue does not pose a hazard to public health or the environment. Based on the risk evaluation, it is proposed that the HTRE-3 Container Storage Unit be closed in its present condition, without further decontamination or removal activities

Investigations on passive containment cooling

International Nuclear Information System (INIS)

Knebel, J.U.; Cheng, X.; Neitzel, H.J.; Erbacher, F.J.; Hofmann, F.

1997-01-01

The composite containment design for advanced LWRs that has been examined under the PASCO project is a promising design concept for purely passive decay heat removal after a severe accident. The passive cooling processes applied are natural convection and radiative heat transfer. Heat transfer through the latter process removes at an emission coefficient of 0.9 about 50% of the total heat removed via the steel containment, and thus is an essential factor. The heat transferring surfaces must have a high emission coefficient. The sump cooling concept examined under the SUCO project achieves a steady, natural convection-driven flow from the heat source to the heat sink. (orig./CB) [de

Cooling device for reactor container

International Nuclear Information System (INIS)

Akiba, Miyuki.

1996-01-01

In a cooling device for a reactor container, a low pressure vessel is connected to an incondensible gas vent tube by way of an opening/closing valve. Upon occurrence of a loss of coolant accident, among steams and incondensible gases contained in the reactor container, steams are cooled and condensed in a heat exchanger. The incondensible gases are at first discharged from the heat exchanger to a suppression pool by way of the incondensible gas vent tube, but subsequently, they are stagnated in the incondensible gas vent tube to hinder heat exchanging and steam cooling and condensing effects in the heat exchanger thereby raising temperature and pressure in the reactor. However, if the opening/closing valve is opened when the incondensible gases are stagnated in the incondensible gas vent tube, since the incondensible gases stagnated in the heat exchanger are sucked and discharged to the low pressure vessel, the performance of the heat exchanger is maintained satisfactorily thereby enabling to suppress elevation of temperature and pressure in the reactor container. (N.H.)

Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

Science.gov (United States)

Soukhanovskii, Vsevolod

2007-11-01

Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required

Heat transfer experiment in a granite formation at Cornwall

International Nuclear Information System (INIS)

Bourke, P.J.; Hodgkinson, D.P.

1985-01-01

An experiment simulating a waste package was started in 1978 in Cornish granite. To obtain measurable temperature rises through large volumes of rock, an electrical heater at 50m depth was run for four years. The heater was then switched off and the cooling was monitored for another year. The results show that most of the heat transfer was by conduction but that some convection occurred and that temperatures can be predicted with some confidence

Results of out-of-pile experiments to investigate the possibilities of cooling a core melt with internal heat production

International Nuclear Information System (INIS)

Fieg, G.

1976-01-01

After serious hypothetical reactor accidents, melted core materials with internal heat production can occur in large quantities. A retention of these molten core masses within the containment must be ensured. The knowledge of the heat transport from volume-heated layers is necessary to clarify this matter. (orig./LH) [de

In-Space technology experiments program. A high efficiency thermal interface (using condensation heat transfer) between a 2-phase fluid loop and heatpipe radiator: Experiment definition phase

Science.gov (United States)

Pohner, John A.; Dempsey, Brian P.; Herold, Leroy M.

1990-01-01

Space Station elements and advanced military spacecraft will require rejection of tens of kilowatts of waste heat. Large space radiators and two-phase heat transport loops will be required. To minimize radiator size and weight, it is critical to minimize the temperature drop between the heat source and sink. Under an Air Force contract, a unique, high-performance heat exchanger is developed for coupling the radiator to the transport loop. Since fluid flow through the heat exchanger is driven by capillary forces which are easily dominated by gravity forces in ground testing, it is necessary to perform microgravity thermal testing to verify the design. This contract consists of an experiment definition phase leading to a preliminary design and cost estimate for a shuttle-based flight experiment of this heat exchanger design. This program will utilize modified hardware from a ground test program for the heat exchanger.

Review of preliminary additional heating experiments in JT-60 (Aug. - Nov., 1986)

International Nuclear Information System (INIS)

1987-03-01

This is a prompt report on preliminary additional heating experiments in JT-60 from August to November in 1986. Neutral beam heating power was raised up to 20 MW in about a month. Plasma stored energy is about 2 MJ and energy confinement time is 0.1 ∼ 0.12 sec with the maximum heating power. The energy confinement time shows L-mode like deterioration with power, while it has little dependence on electron density. The maximum ion temperature of ∼ 7 keV and electron temperature of 4.5 keV were obtained at relatively low electron density (n-bar e = 2 - 3 x 10 19 m -3 ). Lower hybrid wave could efficiently drive plasma current up to 1.7 MA with 1.2 MW LH power. The current drive efficiency is 1 ∼ 1.7 in ohmically heated plasmas and 2 ∼ 2.8 in NB heated plasmas. Futhermore the energy confinement was improved when neutral beam was injected into entirely current driven discharges of 1 MA by LH in contrast to inductively driven target plasmas. Similar improvement in energy confinement was observed during combined heating with NB and ion cyclotron wave. (author)

Thermal-Hydraulic Analysis of the Nuclear Power Engineering Corporation Containment Experiments with GOTHIC

International Nuclear Information System (INIS)

Wiles, Lawrence E.; George, Thomas L.

2003-01-01

GOTHIC version 7.0 was used to model five tests that were conducted in the Nuclear Power Engineering Corporation facility in Japan. The tests involved steam and helium injection into a preheated, spray-moderated, 1/4-scale model of a pressurized water reactor dry containment. Comparison of GOTHIC predictions to measured data for pressure, vapor temperatures, structure surface temperatures, and helium concentrations provided the opportunity to evaluate methods for modeling gas dispersion, drop heat and mass transfer, and surface heat transfer.The test facility includes three floors. The lower two floors are partitioned into a variety of rooms that simulate the lower regions of the modeled containment. On the upper floor, rooms that simulate the steam generator enclosures and the pressurizer enclosure extend into the dome, which represents about two-thirds of the total volume of the containment.The GOTHIC model was defined with 30 control volumes using a mix of lumped parameter volumes and subdivided volumes that employ a three-dimensional mesh. Each volume included several thermal conductors to model the various structures. More than 100 flow paths were used to model the hydraulic connections.Comparison of predictions to data showed that enhanced grid resolution in the vicinity of the steam-helium release point served to limit dispersion of the steam-helium plume. The data comparisons also suggested that spray effectiveness was reduced by drop impact with the containment wall and by the high drop concentration. The data comparisons further suggested that the presence of condensation, sprays, splashing, and other wetting mechanisms should be considered to obtain a reasonable estimate of the effect of liquid films on the structure surfaces

Heat transfer between immiscible liquids enhanced by gas bubbling

International Nuclear Information System (INIS)

Greene, G.A.; Schwarz, C.E.; Klages, J.; Klein, J.

1982-08-01

The phenomena of core-concrete interactions impact upon containment integrity of light water reactors (LWR) following postulated complete meltdown of the core by containment pressurization, production of combustible gases, and basemat penetration. Experiments have been performed with non-reactor materials to investigate one aspect of this problem, heat transfer between overlying immiscible liquids whose interface is disturbed by a transverse non-condensable gas flux emanating from below. Hydrodynamic studies have been performed to test a criterion for onset of entrainment due to bubbling through the interface and subsequent heat transfer studies were performed to assess the effect of bubbling on interfacial heat transfer rates, both with and without bubble induced entrainment. Non-entraining interfacial heat transfer data with mercury-water/oil fluid pairs were observed to be bounded from below within a factor of two to three by the Szekeley surface renewal heat transfer model. However heat transfer data for fluid pairs which are found to entrain (water-oil), believed to be characteristic of molten reactor core-concrete conditions, were measured to be up to two orders of magnitude greater than surface renewal predictions and are calculated by a simple entrainment heat transfer model

Radioactive materials transporting container and vehicles

International Nuclear Information System (INIS)

Reese, S.L.

1980-01-01

A container and vehicle therefor for transporting radioactive materials is provided. The container utilizes a removable system of heat conducting fins made of a light weight highly heat conductive metal, such as aluminum or aluminum alloys. This permits a substantial reduction in the weight of the container during transport, increases the heat dissipation capability of the container and substantially reduces the scrubbing operation after loading and before unloading the radioactive material from the container. The vehicle utilizes only a pair of horizontal side beams interconnecting a pair of yoke members to support the container and provide the necessary strength and safety with a minimum of weight

Major results of the electron cyclotron heating experiment in the PDX tokamak

International Nuclear Information System (INIS)

Hsuan, H.; Bol, K.; Bowen, N.

1984-07-01

Electron Cyclotron Heating (ECH) experiments on PDX have been carried out with two 60 GHz pulsed gyrotrons each yielding up to approximately 100 kW. The ECH system used two waveguide runs each about 30 meters long. One run included 5 bends and the other, 7 bends. Predetermined waveguide modes were transmitted. The electron cyclotron waves were launched in narrow beams from both the high field and the low field sides of the plasma torus. The major new physics results are: (1) efficient central electron heating for both ohmic and neutral beam heated target plasmas; (2) alteration of MHD behavior using ECH; (3) identification of the trapped electron population with ECH; and (4) signature of velocity-space time evolution during ECH. In the best heating results obtained, Thomson scattering data indicated a central temperature increase from less than or equal to 1.5 keV to greater than or equal to 2.5 keV. This occurred with an average density of about 10 13 cm -3 and approximately 80 kW outside-launch ordinary-mode heating

Analysis of the temperature and pore water pressure field in the TED heating experiment

International Nuclear Information System (INIS)

Garitte, B.; Vaunat, J.; Gens, A.; Conil, N.; Armand, G

2012-01-01

Document available in extended abstract form only. The TED experiment is a heating experiment in Callovo-Oxfordian Clay performed in the Meuse Haute Marne underground research laboratory. It was set up by Andra to confirm the results obtained in a previous thermal experiment (TER). The main objectives are: characterisation of the thermal properties of intact argillite, identification of the Thermo-Hydro-Mechanical (THM) coupling parameters, improvement of the understanding of THM behaviour and comparison of parameters measured in-situ and in the laboratory. The three heaters in the TED experiment were installed parallel at a distance of about 2.7 m to reproduce as closely as possible the real storage case (HA cells). This arrangement should allow the verification of: i) the superposition of several temperature fields generated by the three heaters and ii) the overpressure generated in the symmetry planes between the three heaters. Three heaters 4 m long and 0.143 m diameter have been installed in parallel boreholes drilled form GED gallery. To limit the influence of the conditions prevailing in the gallery, a distance of 12 m have been left between the GED wall and the closest extremity of the heaters. Rock has been intensively instrumented within a zone of 15 m around the heaters. There are 108 thermal sensors, 10 pore pressure sensors, 2 extensometers/inclinometers, placed in 13 different small diameter boreholes (56 mm). 11 boreholes were filled with bentonite (those containing thermal and pore pressure sensors) and 2 of them left open (those equipped with extensometer/inclinometers). First heating started on 22 January 2010 at the central heater and consists of three steps: 120 days at 180 W, 150 days at 300 W and one last period at 600 W. The two lateral heaters have been switched on after 14 months following the same heating sequence. The concept and the results of the TED experiment are fully described in a companion paper. Concerning heat transport

Application of heat pipes in nuclear reactors for passive heat removal

Energy Technology Data Exchange (ETDEWEB)

Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-07-01

This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

Controller tuning of district heating networks using experiment design techniques

International Nuclear Information System (INIS)

Dobos, Laszlo; Abonyi, Janos

2011-01-01

There are various governmental policies aimed at reducing the dependence on fossil fuels for space heating and the reduction in its associated emission of greenhouse gases. DHNs (District heating networks) could provide an efficient method for house and space heating by utilizing residual industrial waste heat. In such systems, heat is produced and/or thermally upgraded in a central plant and then distributed to the end users through a pipeline network. The control strategies of these networks are rather difficult thanks to the non-linearity of the system and the strong interconnection between the controlled variables. That is why a NMPC (non-linear model predictive controller) could be applied to be able to fulfill the heat demand of the consumers. The main objective of this paper is to propose a tuning method for the applied NMPC to fulfill the control goal as soon as possible. The performance of the controller is characterized by an economic cost function based on pre-defined operation ranges. A methodology from the field of experiment design is applied to tune the model predictive controller to reach the best performance. The efficiency of the proposed methodology is proven throughout a case study of a simulated NMPC controlled DHN. -- Highlights: → To improve the energetic and economic efficiency of a DHN an appropriate control system is necessary. → The time consumption of transitions can be shortened with the proper control system. → A NLMPC is proposed as control system. → The NLMPC is tuned by utilization of simplex methodology, using an economic oriented cost function. → The proposed NLMPC needs a detailed model of the DHN based on the physical description.

Modelling of the behaviour of a UF_6 container in a fire

International Nuclear Information System (INIS)

Pinton, Eric

1996-01-01

This thesis is justified by the safety needs about storage and transport of UF_6 containers. To define their behaviour under fire conditions, a modelling was developed. Before tackling the numerical modelling, a phenomenological interpretation with experimental results of containers inside a furnace (800 C) during a fixed period was carried out. The internal heat transfers were considerably improved with these results. The 2D elaborated model takes into account most of the physical phenomena encountered in this type of situation (boiling, evaporation, condensation, radiant heat transfers through an absorbing gas, convection, pressurisation, thermal contact resistance, UF_6 expansion, solid core sinking in the liquid, elastic and plastic deformations of the steel container). This model was successfully confronted with experiments. (author) [fr

Modelling hot electron generation in short pulse target heating experiments

Directory of Open Access Journals (Sweden)

Sircombe N.J.

2013-11-01

Full Text Available Target heating experiments planned for the Orion laser facility, and electron beam driven fast ignition schemes, rely on the interaction of a short pulse high intensity laser with dense material to generate a flux of energetic electrons. It is essential that the characteristics of this electron source are well known in order to inform transport models in radiation hydrodynamics codes and allow effective evaluation of experimental results and forward modelling of future campaigns. We present results obtained with the particle in cell (PIC code EPOCH for realistic target and laser parameters, including first and second harmonic light. The hot electron distributions are characterised and their implications for onward transport and target heating are considered with the aid of the Monte-Carlo transport code THOR.

A two-parameter nondiffusive heat conduction model for data analysis in pump-probe experiments

Science.gov (United States)

Ma, Yanbao

2014-12-01

Nondiffusive heat transfer has attracted intensive research interests in last 50 years because of its importance in fundamental physics and engineering applications. It has unique features that cannot be described by the Fourier law. However, current studies of nondiffusive heat transfer still focus on studying the effective thermal conductivity within the framework of the Fourier law due to a lack of a well-accepted replacement. Here, we show that nondiffusive heat conduction can be characterized by two inherent material properties: a diffusive thermal conductivity and a ballistic transport length. We also present a two-parameter heat conduction model and demonstrate its validity in different pump-probe experiments. This model not only offers new insights of nondiffusive heat conduction but also opens up new avenues for the studies of nondiffusive heat transfer outside the framework of the Fourier law.

A thin gold coated hydrogen heat pipe -cryogenic target for external experiments at cosy

International Nuclear Information System (INIS)

Abdel-Bary, M.; Abdel-Samad, S.; Elawadi, G.A.; Kilian, K.; Ritman, J.

2008-01-01

A gravity assisted Gold Coated Heat Pipe (GCHP) with 5-mm diameter has been developed and tested to cool a liquid hydrogen target for external beam experiments at COSY. The need for a narrow target diameter leads us to study the effect of reducing the heat pipe diameter to 5 mm instead of 7 mm, to study the effect of coating the external surface of the heat pipe by a polished gold layer (to decrease the radiation heat load), and to study the effect of using the heat pipe without using 20 layers super isolation around it (aluminized Mylar foil) to keep the target diameter as small as possible. The developed gold coated heat pipe was tested with 20 layers of super isolation and without. The operating characteristics for both conditions were compared to show the advantages and disadvantages

Lower hybrid heating data on the Wega experiment revisited using ion stochastic heating and electron Landau damping theories

International Nuclear Information System (INIS)

Gormezano, C.; Hess, W.; Ichtchenko, G.

1980-07-01

The already obtained data on the Wega Tokamak by lower hybrid heating (f=500 MHz - Psub(HF)=130 KW) are revisited in the light of recent theories on ion stochastic heating and quasi-linear electron Landau damping. It is possible to correctly estimate with these theories the fast ion mean energy, the H.F. power density coupled to the ions and that coupled to the electrons. The values of the parallel index of refraction, Nsub(//), which are necessary to obtain a good quantitative agreement between experiment and theoretical estimates, are the same for the ions and for the electrons, even though at higher values than expected

Use of integral experiments to improve neutron propagation and gamma heating calculations

International Nuclear Information System (INIS)

Oceraies, Y.; Caumette, P.; Devillers, C.; Bussac, J.

1979-01-01

1) The studies to define and improve the accuracies of neutron propagation and gamma heating calculations from integral experiments are encompassed in the field of the fast reactor physics program at CEA. 2) A systematic analysis of neutron propagation in Fe-Na clean media, with variable volumic composition between 0 and 100% in sodium, has been performed on the HARMONIE source reactor. Gamma heating traverses in the core, the blankets and several control rods, have been measured in the R Z core program at MASURCA. The experimental techniques, the accuracies and the results obtained are given. The approximations of the calculational methods used to analyse these experiments and to predict the corresponding design parameters are also described. 3) Particular emphasis is given to the methods planned to improve fundamental data used in neutron propagation calculations, using the discrepancies observed between measured and calculated results in clean integral experiments. One of these approaches, similar to the techniques used in core physics, relies upon sensitivity studies and eventually on adjustment techniques applied to neutron propagation. (author)

Heat Transfer and Latent Heat Storage in Inorganic Molten Salts for Concentrating Solar Power Plants

Energy Technology Data Exchange (ETDEWEB)

Mathur, Anoop [Terrafore Inc.

2013-08-14

A key technological issue facing the success of future Concentrating Solar Thermal Power (CSP) plants is creating an economical Thermal Energy Storage (TES) system. Current TES systems use either sensible heat in fluids such as oil, or molten salts, or use thermal stratification in a dual-media consisting of a solid and a heat-transfer fluid. However, utilizing the heat of fusion in inorganic molten salt mixtures in addition to sensible heat , as in a Phase change material (PCM)-based TES, can significantly increase the energy density of storage requiring less salt and smaller containers. A major issue that is preventing the commercial use of PCM-based TES is that it is difficult to discharge the latent heat stored in the PCM melt. This is because when heat is extracted, the melt solidifies onto the heat exchanger surface decreasing the heat transfer. Even a few millimeters of thickness of solid material on heat transfer surface results in a large drop in heat transfer due to the low thermal conductivity of solid PCM. Thus, to maintain the desired heat rate, the heat exchange area must be large which increases cost. This project demonstrated that the heat transfer coefficient can be increase ten-fold by using forced convection by pumping a hyper-eutectic salt mixture over specially coated heat exchanger tubes. However,only 15% of the latent heat is used against a goal of 40% resulting in a projected cost savings of only 17% against a goal of 30%. Based on the failure mode effect analysis and experience with pumping salt at near freezing point significant care must be used during operation which can increase the operating costs. Therefore, we conclude the savings are marginal to justify using this concept for PCM-TES over a two-tank TES. The report documents the specialty coatings, the composition and morphology of hypereutectic salt mixtures and the results from the experiment conducted with the active heat exchanger along with the lessons learnt during

Facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies

International Nuclear Information System (INIS)

Hennings, U.

1987-01-01

Patent for facility for the storage of spent, heat-emitting and container-enclosed nuclear reactor fuel assemblies, which are arranged within a building in a horizontal position and are cooled by a gas stream, whereby the building has a storage and a loading zone, characterized by the fact that pallet trucks arranged one above the other in a row and such that an interspace is left for the receiving positions for the containers, the the pallet trucks can be moved along rails that extend between two side walls arranged opposite to one another in the storage zone, that the storage zone can be loaded and unloaded by opening located in these two side walls, and that the gas stream only circulates within the building

Sargent and Lundy containment tests revisited

International Nuclear Information System (INIS)

Henry, Robert E.; Hammersley, Robert J.

2005-01-01

The pressurization experiments performed in the intermediate scale Sargent and Lundy containment test facility provide numerous insights into the dominant heat and mass transfer processes under design basis accident conditions similar to a large break Loss of Coolant Accident (LOCA). These experiments were the first integral tests to examine the containment response to a dynamic blowdown from the Reactor Coolant System (RCS). Measurements included the blowdown rate of the simulated Reactor Pressure Vessel (RPV), the pressure in containment as well as the containment temperatures in the top and bottom of the containment vessel. Furthermore, various experiments were performed with the blowdown location changed from the vessel bottom to the lower third of the vessel, the upper third of the vessel and near the top of the RPV to examine the influence of different types of break elevations, i.e. different characterizations of the exhausting steam-water mixture. Perhaps the most insightful set of measurements from these experiments were those that varied the cold water mass initially resident in the bottom of the simulated containment vessel. The role of this water as a function of its initial mass and the break location showed substantial influence of this water if the blowdown location provided sufficient energy to disperse this cold water into the containment building atmosphere. This is demonstrated in Figure 1 taken from Kolflat, 1960. All of these are relevant to an understanding of the dominant physical processes for this type of postulated accident condition. As such, it is important that all of these insights are retained and used in models for the containment building thermal-hydraulic response under accident conditions. Reference: Kolflat, A., 1960, 'Resulting of 1959 Nuclear Power Plant Containment Test', Sargent and Lundy Report SL-1800; Kolflat, A. and Chittenden, W. A., 1957, 'A New Approach to the Design of Containment Shells for Atomic Power Plants

Experience with high heat flux components in large tokamaks

International Nuclear Information System (INIS)

Chappuis, P.; Dietz, K.J.; Ulrickson, M.

1991-01-01

The large present day tokamaks. i.e.JET, TFTR, JT-60, DIII-D and Tore Supra are machines capable of sustaining plasma currents of several million amperes. Pulse durations range from a few seconds up to a minute. These large machines have been in operation for several years and there exists wide experience with materials for plasma facing components. Bare and coated metals, bare and coated graphites and beryllium were used for walls, limiters and divertors. High heat flux components are mainly radiation cooled, but stationary cooling for long pulse duration is also employed. This paper summarizes the experience gained in the large machines with respect to material selection, component design, problem areas, and plasma performance. 2 tabs., 26 figs., 50 refs

An experiment in heat conduction using hollow cylinders

Energy Technology Data Exchange (ETDEWEB)

Ortuno, M; Marquez, A; Gallego, S; Neipp, C; Belendez, A, E-mail: a.belendez@ua.es [Departamento de Fisica, IngenierIa de Sistemas y TeorIa de la Senal, Universidad de Alicante, Apartado 99, E-03080 Alicante (Spain)

2011-07-15

An experimental apparatus was designed and built to allow students to carry out heat conduction experiments in hollow cylinders made of different materials, as well as to determine the thermal conductivity of these materials. The evolution of the temperature difference between the inner and outer walls of the cylinder as a function of time is analysed, and when the process reaches the steady state regime the thermal conductivity can be easily calculated. Several materials such as wood, plastic and metals are considered and the values of their thermal conductivities, obtained experimentally, are compared with those given in the reference list.

Pilot plant experiments for baking of anode blocks in electrically heated ovens

Energy Technology Data Exchange (ETDEWEB)

Grjotheim, K. (Oslo Univ. (Norway). Dept. of Chemistry); Kvande, H. (Hydro Aluminium AS, Stabekk (Norway)); Naixiang, F.; Shiheng, Z.; An, L.; Guangxia, H. (Northeast Univ. of Technology, Shenyang, LN (China). Dept. of Non-Ferrous Metallurgy)

1990-04-01

Pilot plant experiments were made to bake anode blocks in electrically heated baking ovens. About 70% of the baked anodes had a specific electrical resistance between 35 and 60 {Omega}xmm{sup 2}xm{sup -1}. About 25% had higher resistances, and these were returned to the baking ovens and used as heating elements in the next baking cycle. The average electrical energy consumption was 1430 kWh per tonne of anodes produced, which is about only 60% of the energy consumption in classical oil or gas-fired baking ovens. (orig.).

Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

International Nuclear Information System (INIS)

Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

2009-01-01

Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

Selection of cooling fluid for an organic Rankine cycle unit recovering heat on a container ship sailing in the Arctic region

DEFF Research Database (Denmark)

Suárez de la Fuente, Santiago; Larsen, Ulrik; Pierobon, Leonardo

2017-01-01

As Arctic sea ice coverage declines it is expected that marine traffic could increase in this northern region due to shorter routes. Navigating in the Arctic offers opportunities and challenges for waste heat recovery systems (WHRS). Lower temperatures require larger heating power on board, hence...... air as coolant. This paper explores the use of two different coolants, air and seawater, for an organic Rankine cycle (ORC) unit using the available waste heat in the scavenge air system of a container ship navigating in Arctic Circle. Using a two-step single objective optimisation process, detailed...

Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

International Nuclear Information System (INIS)

Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

2004-01-01

Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

A thin gold coated hydrogen heat pipe-cryogenic target for external experiments at COSY

Science.gov (United States)

Abdel-Bary, M.; Abdel-Samad, S.; Elawadi, G. A.; Kilian, K.; Ritman, J.

2009-05-01

A gravity assisted Gold coated heat pipe (GCHP) with 5-mm diameter has been developed and tested to cool a liquid hydrogen target for external beam experiments at COSY. The need for a narrow target diameter leads us to study the effect of reducing the heat pipe diameter to 5 mm instead of 7 mm, to study the effect of coating the external surface of the heat pipe by a shiny gold layer (to decrease the radiation heat load), and to study the effect of using the heat pipe without using 20 layers of' super-insulation around it (aluminized Mylar foil) to keep the target diameter as small as possible. The developed gold coated heat pipe was tested with 20 layers of super-insulation (WI) and without super-insulation (WOI). The operating characteristics for both conditions were compared to show the advantages and disadvantages.

Analysis of direct containment heating in Ling'ao-2 nuclear power plant model With FCI-CMFD software

International Nuclear Information System (INIS)

Wang Xi; Yang Yanhua; Huang Xi

2010-01-01

Ling'ao 2 Nuclear Power Plant (NPP) model was established in the paper. Three dimensional numerical calculation code MC3D was used for analyzing the potential direct containment heating (DCH) accident in the model. MC3D is a special software for fuel coolant interaction (FCI) analysis. In order to predict the accident accurately, initial conditions provided by the SBO accident calculation and the geometric model of Ling'ao 2 NPP were combined to simulate the process of the accident. The calculation gives temperature field, droplet Janume fraction field, velocity field, and the pressure variation in the containment in the case of accident. The result shows that DCH will cause the pressure rising rapidly in the containment and in some local areas temperature rises too. (authors)

Design and operation of a transistorised bridge-type detector for burn-out in boiling heat transfer experiments

Energy Technology Data Exchange (ETDEWEB)

Salt, K J; Wintle, C A [Reactor Development Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1964-04-15

Theoretical, technical and operational details of an instrument to protect heater rods in heat transfer burn-out experiments from actual rupture under electrical heating, by a Wheatstone bridge method. (author)

Design and operation of a transistorised bridge-type detector for burn-out in boiling heat transfer experiments

International Nuclear Information System (INIS)

Salt, K.J.; Wintle, C.A.

1964-04-01

Theoretical, technical and operational details of an instrument to protect heater rods in heat transfer burn-out experiments from actual rupture under electrical heating, by a Wheatstone bridge method. (author)

NORDA contribution to the in-situ heat transfer experiment (ISHTE): FY84 annual report

International Nuclear Information System (INIS)

Valent, P.J.; Bennett, R.H.; Li, H.; Burns, J.T.

1986-01-01

The Subseabed Disposal Program (SDP) of the DOE, managed by Sandia National Laboratories, Albuquerque (SNLA), is studying the feasibility of disposing of high-level radioactive wastes by burial in fine-grained deep-sea sediments. The thermo-mechanical response of these sediments to the thermal gradient and temperatures generated by the decaying radionucleides in a buried waste container is being determined by the SDP-supported In Situ Heat Transfer Experiment (ISHTE). The Naval Ocean Research and Development Activity (NORDA) is responsible for the development and fielding of piezometer probes for measuring the pore water pressure gradients induced by the thermal gradient in the sediment. Pore pressure gradients measured in ISHTE will permit validation of theoretical models predicting the rate of radionucleide leakage from a buried waste container to the overlying seawater column. This report of describes the results of a laboratory simulation of ISHTE, conducted at SNLA, in which sediment cracking due to probe insertion was determined to not be a problem to the experiment. Specialized equipment developed for ISHTE, in particular a pressure transducer calibrator for ambient pressures to 69 MPa (10,000 psi), is described. Preliminary results gleaned from excess pore pressure data obtained during the ISHTE component test cruise, Sept 84, confirm that sediment cracking due to probe insertion will be slight and will have an insignificant influence on the measured excess pore pressure dissipation rates. Excess preenerated by insertion of the piezometer probes vary by a factor of two. No significant faults in the NORDA piezometer system were detected either in the laboratory stimulation or in the test in 5800 m water depth north of Hawaii

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Thulium-170 heat source

Science.gov (United States)

Walter, Carl E.; Van Konynenburg, Richard; VanSant, James H.

1992-01-01

An isotopic heat source is formed using stacks of thin individual layers of a refractory isotopic fuel, preferably thulium oxide, alternating with layers of a low atomic weight diluent, preferably graphite. The graphite serves several functions: to act as a moderator during neutron irradiation, to minimize bremsstrahlung radiation, and to facilitate heat transfer. The fuel stacks are inserted into a heat block, which is encased in a sealed, insulated and shielded structural container. Heat pipes are inserted in the heat block and contain a working fluid. The heat pipe working fluid transfers heat from the heat block to a heat exchanger for power conversion. Single phase gas pressure controls the flow of the working fluid for maximum heat exchange and to provide passive cooling.

Analysis of CSNI benchmark test on containment using the code CONTRAN

International Nuclear Information System (INIS)

Haware, S.K.; Ghosh, A.K.; Raj, V.V.; Kakodkar, A.

1994-01-01

A programme of experimental as well as analytical studies on the behaviour of nuclear reactor containment is being actively pursued. A large number ol' experiments on pressure and temperature transients have been carried out on a one-tenth scale model vapour suppression pool containment experimental facility, simulating the 220 MWe Indian Pressurised Heavy Water Reactors. A programme of development of computer codes is underway to enable prediction of containment behaviour under accident conditions. This includes codes for pressure and temperature transients, hydrogen behaviour, aerosol behaviour etc. As a part of this ongoing work, the code CONTRAN (CONtainment TRansient ANalysis) has been developed for predicting the thermal hydraulic transients in a multicompartment containment. For the assessment of the hydrogen behaviour, the models for hydrogen transportation in a multicompartment configuration and hydrogen combustion have been incorporated in the code CONTRAN. The code also has models for the heat and mass transfer due to condensation and convection heat transfer. The structural heat transfer is modeled using the one-dimensional transient heat conduction equation. Extensive validation exercises have been carried out with the code CONTRAN. The code CONTRAN has been successfully used for the analysis of the benchmark test devised by Committee on the Safety of Nuclear Installations (CSNI) of the Organisation for Economic Cooperation and Development (OECD), to test the numerical accuracy and convergence errors in the computation of mass and energy conservation for the fluid and in the computation of heat conduction in structural walls. The salient features of the code CONTRAN, description of the CSNI benchmark test and a comparison of the CONTRAN predictions with the benchmark test results are presented and discussed in the paper. (author)

Heat recovery apparatus

International Nuclear Information System (INIS)

McFarland, I.

1987-01-01

Heat transfer is a living science and technical advances are constantly being made. However, in many cases, progress is limited by the equipment that is available on the market, rather than by knowledge of the heat transfer process. A case in point is the design of economizers: in such equipment a small quantity of water (with a relatively good heat transfer coefficient) is heated by a large quantity of low-pressure gas (with an inherently low heat transfer coefficient). As a first step in design finned tubing is used to lessen the discrepancy in coefficients. From this point, it becomes apparent that the equipment consists of a small number of tubes (to maintain good velocity on the water side) of considerable length (to provide sufficient area). In the process industries the base pressure, though low, may be in the region of 0.5 bar, and there is no convenient flue in which to place the heat recovery coil. It is therefore contained in a flat-sided enclosure, which is ill-fitted to pressure containment and is therefore reinforced with a plethora of structural sections. Such inelegant construction is quite common in North America; in Europe, cylindrical containments of vast size have been supplied for the same purposes. The real shortcoming is a successful marriage of different disciplines to produce reliable and efficient heat transfer equipment suitably contained

The assessment of containment codes by experiments simulating severe accident scenarios

International Nuclear Information System (INIS)

Karwat, H.

1992-01-01

Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

Disposal/storage container development experience

International Nuclear Information System (INIS)

Morrow, R.W. Jr.; Van Hoesen, S.D.; Fowler, E.; Barreira, D.G.; Emmett, R.W.

1988-01-01

Developmental work is currently underway at the Oak Ridge National Laboratory to design and manufacture a radioactive waste container suitable for both storage and disposal of radioactive wastes. The container is designed to fulfill the Department of Energy and Nuclear Regulatory Commission requirements for on-site storage, as well as the Nuclear Regulatory Commission's requirements for high integrity containers. The project also involves meeting the strict design and manufacturing ANSI/ASME NQA-1 guidelines. Special provisions of the container include a double containment system, with the inner barrier being corrosion resistant, the capability to monitor the internal cavity of the container, and off-gas venting capability. Further, yet related developmental work includes evaluating the cask for other varied uses, such as a processing cask, an ALARA shield, and even the possibility of Department of Transportation approval for an over-the-road transport cask

Critical heat flux experiments for high conversion light water reactor, (3)

International Nuclear Information System (INIS)

Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

1990-03-01

As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

Laboratory experiments on heat-drive two-phase flows in natural and artificial rock fractures

International Nuclear Information System (INIS)

Kneafsey, Timothy J.; Pruess, Karsten

1998-01-01

Water flow in partially saturated fractures under thermal drive may lead to fast flow along preferential localized pathways and heat pipe conditions. At the potential high-level nuclear waste repository at Yucca Mountain, water flowing in fast pathways may ultimately contact waste packages and transport radionuclides to the accessible environment. Sixteen experiments were conducted to visualize heat-driven liquid flow in fracture models that included (1) assemblies of roughened glass plates, (2) epoxy replicas of rock fractures, and (3) a fractured specimen of Topopah Spring tuff. Continuous rivulet flow was observed for high liquid flow rates, intermittent rivulet flow and drop flow for intermediate flow rates, and film flow for lower flow rates and wide apertures. Heat pipe conditions (vapor-liquid counterflow with phase change) were identified in five of the seven experiments in which spatially resolved thermal monitoring was performed but not when vapor-liquid counterflow was hindered by very narrow apertures and when an inadequate working fluid volume was used

Large Eddy/Reynolds-Averaged Navier-Stokes Simulations of CUBRC Base Heating Experiments

Science.gov (United States)

Salazar, Giovanni; Edwards, Jack R.; Amar, Adam J.

2012-01-01

ven with great advances in computational techniques and computing power during recent decades, the modeling of unsteady separated flows, such as those encountered in the wake of a re-entry vehicle, continues to be one of the most challenging problems in CFD. Of most interest to the aerothermodynamics community is accurately predicting transient heating loads on the base of a blunt body, which would result in reduced uncertainties and safety margins when designing a re-entry vehicle. However, the prediction of heat transfer can vary widely depending on the turbulence model employed. Therefore, selecting a turbulence model which realistically captures as much of the flow physics as possible will result in improved results. Reynolds Averaged Navier Stokes (RANS) models have become increasingly popular due to their good performance with attached flows, and the relatively quick turnaround time to obtain results. However, RANS methods cannot accurately simulate unsteady separated wake flows, and running direct numerical simulation (DNS) on such complex flows is currently too computationally expensive. Large Eddy Simulation (LES) techniques allow for the computation of the large eddies, which contain most of the Reynolds stress, while modeling the smaller (subgrid) eddies. This results in models which are more computationally expensive than RANS methods, but not as prohibitive as DNS. By complimenting an LES approach with a RANS model, a hybrid LES/RANS method resolves the larger turbulent scales away from surfaces with LES, and switches to a RANS model inside boundary layers. As pointed out by Bertin et al., this type of hybrid approach has shown a lot of promise for predicting turbulent flows, but work is needed to verify that these models work well in hypersonic flows. The very limited amounts of flight and experimental data available presents an additional challenge for researchers. Recently, a joint study by NASA and CUBRC has focused on collecting heat transfer data

Comparing the analgesic effect of heat patch containing iron chip and ibuprofen for primary dysmenorrhea: a randomized controlled trial

Directory of Open Access Journals (Sweden)

Navvabi Rigi Shahindokht

2012-08-01

Full Text Available Abstract Background Primary dysmenorrhea is a common and sometimes disabling condition. In recent years, some studies aimed to improve the treatment of dysmenorrhea, and therefore, introduced several therapeutic measures. This study was designed to compare the analgesic effect of iron chip containing heat wrap with ibuprofen for the treatment of primary dysmenorrhea. Methods In this randomized (IRCT201107187038N2 controlled trial, 147 students (18–30 years old with the diagnosis of primary dysmenorrhea were enrolled considering the CONSORT guideline. Screening for primary dysmenorrhea was done by a two-question screening tool. The participants were randomly assigned into one of the intervention groups (heat Patch and ibuprofen. Data regarding the severity and emotional impact of the pain were recorded by a shortened version of McGill Pain Questionnaire (SF-MPQ. Student's t test was used for statistical analysis. Results The maximum and minimum pain severities were observed at 2 and 24 hours in both groups. The severity of sensual pain at 8, 12, and 24 hours was non-significantly less in the heat Patch group. There was also no significant difference between the groups regarding the emotional impact of pain at the first 2, 4, 8, 12 and 12 hours of menstruation. Conclusions Heat patch containing Iron chip has comparable analgesic effects to ibuprofen and can possibly be used for primary dysmenorrhea. Trial registration IRCT201107187038N2

Natural convection heat transfer in SIGMA experiment

International Nuclear Information System (INIS)

Lee, Seung Dong; Lee, Gang Hee; Suh, Kune Yull

2004-01-01

A loss-of-coolant accident (LOCA) results in core melt formation and relocation at various locations within the reactor core over a considerable period of time. If there is no effective cooling mechanism, the core debris may heat up and commence natural circulation. The high temperature pool of molten core material will threaten the structural integrity of the reactor vessel. The extent and urgency of this threat depend primarily upon the intensity of the internal heat sources and upon the consequent distribution of the heat fluxes on the vessel walls in contact with the molten core material pools. In such a steady molten pool convection state, the thermal loads against the vessel would be determined by the in-vessel heat transfer distribution involving convective and conductive heat transfer from the decay-heated core material pool to the lower head wall in contact with the core material. In this study, upward and downward heat transfer fraction ratio is focused on

Numerical Heat Transfer Studies of a Latent Heat Storage System Containing Nano-Enhanced Phase Change Material

Directory of Open Access Journals (Sweden)

S F Hosseinizadeh

2011-01-01

Full Text Available The heat transfer enhancement in the latent heat thermal energy storage system through dispersion of nanoparticle is reported. The resulting nanoparticle-enhanced phase change materials (NEPCM exhibit enhanced thermal conductivity in comparison to the base material. The effects of nanoparticle volume fraction and some other parameters such as natural convection are studied in terms of solid fraction and the shape of the solid-liquid phase front. It has been found that higher nanoparticle volume fraction result in a larger solid fraction. The present results illustrate that the suspended nanoparticles substantially increase the heat transfer rate and also the nanofluid heat transfer rate increases with an increase in the nanoparticles volume fraction. The increase of the heat release rate of the NEPCM shows its great potential for diverse thermal energy storage application.

Preliminary results from initial in-pile debris bed experiments

International Nuclear Information System (INIS)

Rivard, J.B.

1977-01-01

An accident in a liquid metal fast breeder reactor (LMFBR) in which molten core material is suddenly quenched with subcooled liquid sodium could result in extensive fragmentation and dispersal of fuel as subcritical beds of frozen particulate debris within the reactor vessel. Since this debris will continue to generate power due to decay of retained fission products, containment of the debris is threatened if the generated heat is not removed. Therefore, the initial safety question is the capacity which debris beds may have for transfer of the decay heat to overlying liquid sodium by natural processes--i.e., without the aid of forced circulation of the coolant. Up to the present time, all experiments on debris bed behavior either have used substitute materials (e.g., sand and water) or have employed actual materials, but atypical heating methods. Increased confidence in the applicability of debris bed simulations is afforded if the heat is generated within the fuel component of the appropriate fast reactor materials. The initial series of in-pile tests reported on herein constitutes the first experiments in which the internal heating mode has been produced in particulate oxide fuel immersed in liquid sodium. Fission heating of the fully-enriched UO 2 in the experiment while it is contained within Sandia Laboratories Annular Core Pulse Reactor (ACPR), operating in its steady-state mode, approximates the decay heating of debris. Preliminary results are discussed

Containment response and radiological release for a TMLB' accident sequence in a large dry containment

International Nuclear Information System (INIS)

Gasser, R.D.; Bieniarz, P.P.; Tills, J.L.

1987-01-01

An analysis has been performed for the Bellefonte Pressurized Water Reactor (PWR) Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis, which include the effects of direct heating on containment loading and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating, which involves more than about 50% of the core, may fail the Bellefonte containment, but natural convection in the Reactor Coolant System (RCS) may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach, due to natural circulation, and after vessel breach, due to reevolution of retained fission products by fission product heating of RCS structures. (orig.)

Helical type vacuum container

International Nuclear Information System (INIS)

Owada, Kimio.

1989-01-01

Helical type vacuum containers in the prior art lack in considerations for thermal expansion stresses to helical coils, and there is a possibility of coil ruptures. The object of the present invention is to avoid the rupture of helical coils wound around the outer surface of a vacuum container against heat expansion if any. That is, bellows or heat expansion absorbing means are disposed to a cross section of a helical type vacuum container. With such a constitution, thermal expansion of helical coils per se due to temperature elevation of the coils during electric supply can be absorbed by expansion of the bellows or absorption of the heat expansion absorbing means. Further, this can be attained by arranging shear pins in the direction perpendicular to the bellows axis so that the bellows are not distorted when the helical coils are wound around the helical type vacuum container. (I.S.)

Heat transfer in a compact heat exchanger containing rectangular channels and using helium gas

Science.gov (United States)

Olson, D. A.

1991-01-01

Development of a National Aerospace Plane (NASP), which will fly at hypersonic speeds, require novel cooling techniques to manage the anticipated high heat fluxes on various components. A compact heat exchanger was constructed consisting of 12 parallel, rectangular channels in a flat piece of commercially pure nickel. The channel specimen was radiatively heated on the top side at heat fluxes of up to 77 W/sq cm, insulated on the back side, and cooled with helium gas flowing in the channels at 3.5 to 7.0 MPa and Reynolds numbers of 1400 to 28,000. The measured friction factor was lower than that of the accepted correlation for fully developed turbulent flow, although the uncertainty was high due to uncertainty in the channel height and a high ratio of dynamic pressure to pressure drop. The measured Nusselt number, when modified to account for differences in fluid properties between the wall and the cooling fluid, agreed with past correlations for fully developed turbulent flow in channels. Flow nonuniformity from channel-to-channel was as high as 12 pct above and 19 pct below the mean flow.

Nuclear heat applications: design aspects and operating experience. Proceedings of four technical meetings held between December 1995 and April 1998

International Nuclear Information System (INIS)

1998-11-01

Proven to be safe, reliable, economic and having minimum impact on the environment, nuclear energy is playing an important role in electricity generation producing 175 of the world's electricity. But since most of the world's energy consumption is in the form of heat the market for nuclear heat has already been recognised. Considering the growing experience in application of power reactors for district heating, industrial processes and water desalination IAEA is periodically reviewing progress and new developments of nuclear heat applications. This proceedings includes the papers presented at the following four meetings: Advisory group meeting and consultancy on experience with nuclear heat applications: district heating, process heat and desalination, 13-15 December 1995 and 7-9 february 1996; Advisory group meeting on technology, design and safety aspects of non-electrical application of nuclear energy, 20-24 October 1997; Advisory group meeting on operational modes of nuclear desalination plants, 3-5 November 1997; Advisory group meeting on materials and equipment for the coupling interfaces of nuclear reactors with desalination and district heating plants, 21-23 April 1998. It is structured according to the subject areas: (1) design an safety aspects of nuclear heat application, (2) operational and material aspects of nuclear heat application and (3) operational experience with nuclear heat application. Each paper is described by a separate abstract

Feasibility study for a postaccident heat removal facility

International Nuclear Information System (INIS)

Barts, E.W.; Apperson, C.E. Jr.; Dunwoody, W.E.; Bennett, J.G.

1978-01-01

An initial feasibility investigation for PAHRTEF, a Postaccident Heat Removal Test Facility, is presented. The facility would provide an experimental capability for PAHR experiments beyond that available in any currently existing or proposed U.S. safety test facility. The facility design presented in this report is based upon the technology developed for the ROVER nuclear rocket propulsion program. The core is a graphite-moderated, helium-cooled, epithermal core with radial reflector control. The PAHR experiments are located just below the reactor containment vessel, very near the bottom of the core. The experiments (up to 55% enriched) are driven and controlled by neutrons leaking axially from the core such that the PAHRTEF core and the experiment form a coupled reactor system. The experiment can be designed so that it is extremely unlikely that the test fuel by itself could form a critical system. The investigation indicates that adequate fission heating of large PAHR experiments could be provided at low driver core power levels. Both the reactor and the experiment handling and examination equipment can use available technology and, whenever possible, existing equipment and buildings

Operation experiences of the JT-60 first walls during high-power additional heating experiments

International Nuclear Information System (INIS)

Takatsu, H.; Ando, T.; Yamamoto, M.; Arai, T.; Kodama, K.; Suzuki, M.; Shimizu, M.

1989-01-01

JT-60 started its operation in May 1985 with TiC-coated molybdenum or Inconel 625 first walls. They provided very clean surfaces as well as superior plasma characteristics during Joule heating discharges. Though 20 μm-thick TiC coatings showed good adhesion characteristics, melting of the TiC coating and also the molybdenum or Inconel 625 substrate was observed at some specific spots, and an influx of heavy metals to the main plasma was inevitable during discharges. Initial results of the additional heating experiments showed degrading effects of locally melted TiC-coated molybdenum or Inconel 625 on plasma operation. Therefore, about a half of the TiC-coated first walls were removed and new graphite first walls were installed during the venting period from April to May 1987. The start-up of the discharge conditioning after installation of a significant number of graphite tiles was very rapid. Flexibility in plasma operation was increased, and JT-60 extended the operation region beyond its original specifications. The graphite first walls of the main chamber performed admirably and maintained their integrity under the conditions of plasma current and additional heating power up to 3.2 MA and 30 MW, respectively. On the other hand, the number of damaged divertor plates was much larger than that expected. The reason of unexpected failure is now under examination. (orig.)

Method of making Tl-Sr-Ca-Cu-oxide superconductors comprising heating at elevated pressures in a sealed container

International Nuclear Information System (INIS)

Lechtev, W.L.; Osofsky, M.S.; Skelton, E.F.; Toth, L.E.

1992-01-01

This patent describes a method of forming a Tl-Sr-Ca-Cu-oxide high T c superconductor. It comprises forming a reaction mixture of the oxides of Sr, Cu, Ca, and Tl in stoichiometric proportions to make a Tl-Sr-Ca-Cu-oxide high T c superconducting compound; compressing the reaction mixture into a hard body; placing the hard body into a container for containing thallium vapor; evacuating and sealing the hard body in the container; heating the hard body and the container at a temperature of about 800 degrees C to about 950 degrees C and under pressure of at least about 30,000 psi until the container metal around the hard body and the oxides of Tl, Sr, Ca, and Cu react to form a superconducting compound; and cooling the superconducting compound to room temperature and returning the superconducting compound to atmospheric pressure

Orion EFT-1 Cavity Heating Tile Experiments and Environment Reconstruction

Science.gov (United States)

Salazar, Giovanni; Amar, Adam; Oliver, Brandon; Hyatt, Andrew; Rezin, Marc

2016-01-01

Developing aerothermodynamic environments for deep cavities, such as those produced by micrometeoroids and orbital debris impacts, poses a great challenge for engineers. In order to assess existing cavity heating models, two one-inch diameter cavities were flown on the Orion Multi-Purpose Crew Vehicle during Exploration Flight Test 1 (EFT1). These cavities were manufactured with depths of 1.0 in and 1.4 in, and they were both instrumented. Instrumentation included surface thermocouples upstream, downstream and within the cavities, and additional thermocouples at the TPS-structure interface. This paper will present the data obtained, and comparisons with computational predictions will be shown. Additionally, the development of a 3D material thermal model will be described, which will be used to account for the three-dimensionality of the problem when interpreting the data. Furthermore, using a multi-dimensional inverse heat conduction approach, a reconstruction of a time- and space-dependent flight heating distribution during EFT1 will be presented. Additional discussions will focus on instrumentation challenges and calibration techniques specific to these experiments. The analysis shown will highlight the accuracies and/or deficiencies of current computational techniques to model cavity flows during hypersonic re-entry.

CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA

International Nuclear Information System (INIS)

Hargroves, D.W.; Metcalfe, L.J.; Cheng, Teh-Chin; Wheat, L.L.; Mings, W.J.

1991-01-01

describing structure behavior are advanced using an implicit technique. The resulting heat transfer rates are used to correct the previous estimates of the water and energy storage in the containment volume, and the containment conditions are obtained by solving for the second time the containment balance equations. The pressure suppression routines use the conditions at the beginning of a time-step to calculate both the initial explosion of water from the vents and the flow through the vents. From the calculated flow rates, mass and energy are removed from the dry well and added to the wet well. 3 - Restrictions on the complexity of the problem - Maxima of: 20 heat conducting structures, 101 mesh points for each structure, 20 regions for each structure, 50 flow elements in one segment of the horizontal vent pressure suppression system, 10 horizontal vents (or branches) in a segment, 50 reductions within an input time-step. CONTEMPT-LT can be used for analyzing the transient containment behavior of boiling-water reactors (BWRs) including Mark I, Mark II, and Mark III systems; pressurized-water reactors (PWRs), and experimental water reactor simulators or related experiments

Pretest parametric calculations for the heated pillar experiment in the WIPP In-Situ Experimental Area

International Nuclear Information System (INIS)

Branstetter, L.J.

1983-03-01

Results are presented for a pretest parametric study of several configurations and heat loads for the heated pillar experiment (Room H) in the Waste Isolation Pilot Plant (WIPP) In Situ Experimental Area. The purpose of this study is to serve as a basis for selection of a final experiment geometry and heat load. The experiment consists of a pillar of undisturbed rock salt surrounded by an excavated annular room. The pillar surface is covered by a blanket heat source which is externally insulated. A total of five thermal and ten structural calculations are described in a four to five year experimental time frame. Results are presented which include relevant temperature-time histories, deformations, rock salt stress component and effective stress profiles, and maximum stresses in anhydrite layers which are in close proximity to the room. Also included are predicted contours of a conservative post-processed measure of potential salt failure. Observed displacement histories are seen to be highly dependent on pillar and room height, but insensitive to other geometrical variations. The use of a tensile cutoff across slidelines is seen to produce more accurate predictions of anhydrite maximum stress, but to have little effect on rock salt stresses. The potential for salt failure is seen to be small in each case for the time frame of interest, and is only seen at longer times in the center of the room floor

Alternative cooling water flow path for RHR heat exchanger and its effect on containment response during extended station blackout for Chinshan BWR-4 plant

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

2016-04-15

Highlights: • Motivating alternative RHR heat exchanger tube-side flow path and determining required capacity. • Calculate NSSS and containment response during 24-h SBO for Chinshan BWR-4 plant. • RETRAN and GOTHIC models are developed for NSSS and containment, respectively. • Safety relief valve blowdown flow and energy to drywell are generated by RETRAN. • Analyses are performed with and without reactor depressurization, respectively. - Abstract: The extended Station Blackout (SBO) of 24 h has been analyzed with respect to the containment response, in particular the suppression pool temperature response, for the Chinshan BWR-4 plant of MARK-I containment. The Chinshan plant, owned by Taiwan Power Company, has twin units with rated core thermal power of 1840 MW each. The analysis is aimed at determining the required alternative cooling water flow capacity for the residual heat removal (RHR) heat exchanger when its tube-side sea water cooling flow path is blocked, due to some reason such as earthquake or tsunami, and is switched to the alternative raw water source. Energy will be dissipated to the suppression pool through safety relief valves (SRVs) of the main steam lines during SBO. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The shaft seals of the two recirculation pumps are conservatively assumed to fail due to loss of seal cooling and a total leakage flow rate of 36 gpm to the drywell is included in the GOTHIC model. Based on the given SRV blowdown conditions, the GOTHIC containment calculation is performed several times, through the adjustment of the heat transfer rate of the RHR heat exchanger, until the criterion that the maximum suppression pool temperature

Review of Current Experience on Intermediate Heat Exchanger (IHX) and A Recommended Code Approach

Energy Technology Data Exchange (ETDEWEB)

Duane Spencer; Kevin McCoy

2010-02-02

The purpose of the ASME/DOE Gen IV Task 7 Part I is to review the current experience on various high temperature reactor intermediate heat exchanger (IHX) concepts. There are several different IHX concepts that could be envisioned for HTR/VHTR applications in a range of temperature from 850C to 950C. The concepts that will be primarily discussed herein are: (1) Tubular Helical Coil Heat Exchanger (THCHE); (2) Plate-Stamped Heat Exchanger (PSHE); (3) Plate-Fin Heat Exchanger (PFHE); and (4) Plate-Machined Heat Exchanger (PMHE). The primary coolant of the NGNP is potentially subject to radioactive contamination by the core as well as contamination from the secondary loop fluid. To isolate the radioactivity to minimize radiation doses to personnel, and protect the primary circuit from contamination, intermediate heat exchangers (IHXs) have been proposed as a means for separating the primary circuit of the NGNP (Next Generation Nuclear Plant) or other process heat application from the remainder of the plant. This task will first review the different concepts of IHX that could be envisioned for HTR/VHTR applications in a range of temperature from 850 to 950 C. This will cover shell-and-tube and compact designs (including the platefin concept). The review will then discuss the maturity of the concepts in terms of design, fabricability and component testing (or feedback from experience when applicable). Particular attention will be paid to the feasibility of developing the IHX concepts for the NGNP with operation expected in 2018-2021. This report will also discuss material candidates for IHX applications and will discuss specific issues that will have to be addressed in the context of the HTR design (thermal aging, corrosion, creep, creep-fatigue, etc). Particular attention will be paid to specific issues associated with operation at the upper end of the creep regime.

Review of Current Experience on Intermediate Heat Exchanger (IHX) and A Recommended Code Approach

International Nuclear Information System (INIS)

Spencer, Duane; McCoy, Kevin

2010-01-01

The purpose of the ASME/DOE Gen IV Task 7 Part I is to review the current experience on various high temperature reactor intermediate heat exchanger (IHX) concepts. There are several different IHX concepts that could be envisioned for HTR/VHTR applications in a range of temperature from 850C to 950C. The concepts that will be primarily discussed herein are: (1) Tubular Helical Coil Heat Exchanger (THCHE); (2) Plate-Stamped Heat Exchanger (PSHE); (3) Plate-Fin Heat Exchanger (PFHE); and (4) Plate-Machined Heat Exchanger (PMHE). The primary coolant of the NGNP is potentially subject to radioactive contamination by the core as well as contamination from the secondary loop fluid. To isolate the radioactivity to minimize radiation doses to personnel, and protect the primary circuit from contamination, intermediate heat exchangers (IHXs) have been proposed as a means for separating the primary circuit of the NGNP (Next Generation Nuclear Plant) or other process heat application from the remainder of the plant. This task will first review the different concepts of IHX that could be envisioned for HTR/VHTR applications in a range of temperature from 850 to 950 C. This will cover shell-and-tube and compact designs (including the platefin concept). The review will then discuss the maturity of the concepts in terms of design, fabricability and component testing (or feedback from experience when applicable). Particular attention will be paid to the feasibility of developing the IHX concepts for the NGNP with operation expected in 2018-2021. This report will also discuss material candidates for IHX applications and will discuss specific issues that will have to be addressed in the context of the HTR design (thermal aging, corrosion, creep, creep-fatigue, etc). Particular attention will be paid to specific issues associated with operation at the upper end of the creep regime.

The effect of heat treatment and irradiation on some physical properties of lithium borate glass containing transition elements

International Nuclear Information System (INIS)

Soliman, A.A.; Aly, S.A.; Frhan, H.; Abo-Zeid, Y.M.

1999-01-01

The effect of introducing iron, nickel and cobalt oxide on some physical properties investigated in this article. The electrical conductivity has a higher value in samples containing 1 mol% transition metal oxides. The conductivity decreased as the content of transition metal oxide increased up to 5 mol% which was due to the change of Fe 2+ to Fe 3+ and increase of Co and Ni ions in octahedral state. The effect of heat treatment on the conductivity measurements shows a decrease in the conductivity values for glass samples with increasing the heat treatment time up to 72 h. This decrease could be attributed to the change in the structure of the glass samples. The investigation of radiation doses with the electrical conductivity concluded that the conductivity increased with increase the irradiation doses. The reason of that may be due to increasing the number of vacancies and vacancy interstitial pairs which are created. The magnetic susceptibility measurements showed an increase in the magnetic susceptibility as Fe 2 O 3 and NiO were increased. While for samples containing CoO the magnetic susceptibility changed due to the change in coordination number of the Co ions. The effect of heat treatment on magnetic susceptibility of the investigated samples concluded that the magnetic susceptibilities have a random behavior with increasing time of heat treatment. By investigating irradiation doses with a magnetic susceptibility it was found that the increase of irradiation dose promotes a tendency to change the magnetic susceptibility values. This change can be related to the presence of structure defects and impurities in the samples before irradiation

Determination of the stochastic layer properties induced by magnetic perturbations via heat pulse experiments at ASDEX upgrade

Directory of Open Access Journals (Sweden)

D. Brida

2017-08-01

Full Text Available A heat pulse experiment was carried out in the tokamak ASDEX Upgrade to estimate the stochastic layer width of a deuterium L-mode discharge with externally applied Magnetic Perturbations. The method relies on the deposition of ECRH pulses in the plasma edge while measuring the divertor target heat flux with high temporal resolution IR thermography and Langmuir probes. The experimental results were compared to simulations of the time dependent heat pulse propagation on a constant plasma background with the EMC3-EIRENE code package, using an ad-hoc screening model. If no screening was taken into account in the simulations a decrease in the characteristic heat pulse propagation time was observed, which shows that the heat transport is enhanced compared to the screened cases. No such enhancement was found in the experiment, indicating strong screening. In further simulations the effect of screening on the target fluxes was investigated for varying densities. For low densities it was found that screening reduces the strike line splitting strongly, while for higher densities no strong strike line splitting was found, independent of the screening degree. For strongly detached L-mode conditions with MPs experiments at AUG indicate that the lobe structures vanish completely.

Fast-wave ICRF minority-regime heating experiments on the Tore Supra tokamak

Energy Technology Data Exchange (ETDEWEB)

Agarici, G; Beaumont, B; Becoulet, A; Kuus, H; Saoutic, B; Martin, G [Association Euratom-CEA, Centre d` Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (FR). Dept. de Recherches sur la Fusion Controlee; Shepard, T D; Haste, G R; Baity, F W [Oak Ridge National Lab., TN (US); Evans, T E [General Atomics, San Diego, CA (US)

1992-12-31

Up to 4 MW of rf power at 57 MHz has been coupled to Ohmic target plasmas during the first ICRF heating experiments on Tore Supra. A total of 12 MW of rf power will ultimately be available from six tetrode amplifiers and will be coupled to the plasmas using three ORNL/CEA-designed resonant double-loop antennas. During these first experiments, two antennas were used, with one or two energized at a time. The antenna loading with plasma was observed to be well over an order of magnitude greater than that without plasma. In addition, one kilo-electron-volt of electron heating, significant minority nonthermal ions, and significant increases in diamagnetic stored energy were observed. A comparison of in-phase and out-of-phase antenna operation showed the same increase in stored energy, less radiated power, and a larger drop in loop voltage for out-of-phase operation. Confinement scaling agrees with the ITER scaling law.

Heat pipe dynamic behavior

Science.gov (United States)

Issacci, F.; Roche, G. L.; Klein, D. B.; Catton, I.

1988-01-01

The vapor flow in a heat pipe was mathematically modeled and the equations governing the transient behavior of the core were solved numerically. The modeled vapor flow is transient, axisymmetric (or two-dimensional) compressible viscous flow in a closed chamber. The two methods of solution are described. The more promising method failed (a mixed Galerkin finite difference method) whereas a more common finite difference method was successful. Preliminary results are presented showing that multi-dimensional flows need to be treated. A model of the liquid phase of a high temperature heat pipe was developed. The model is intended to be coupled to a vapor phase model for the complete solution of the heat pipe problem. The mathematical equations are formulated consistent with physical processes while allowing a computationally efficient solution. The model simulates time dependent characteristics of concern to the liquid phase including input phase change, output heat fluxes, liquid temperatures, container temperatures, liquid velocities, and liquid pressure. Preliminary results were obtained for two heat pipe startup cases. The heat pipe studied used lithium as the working fluid and an annular wick configuration. Recommendations for implementation based on the results obtained are presented. Experimental studies were initiated using a rectangular heat pipe. Both twin beam laser holography and laser Doppler anemometry were investigated. Preliminary experiments were completed and results are reported.

Heat transfer and friction characteristics of rotor-assembled strand heat exchanger studied by uniform design experiment

Directory of Open Access Journals (Sweden)

Yan Wei

2015-10-01

Full Text Available The uniform distribution and experimental design is employed to study the thermo-hydraulic characteristics of a heat exchanger, which consists of the rotor-assembled strands mounted in circular smooth tubes. The uniform distribution and experimental design parameters include multiple rotor parameters such as rotor diameters, rotor lead, and height of blade, with the aim of studying their influence on the PEC, that is, ( ( Nu z / Nu g / ( f g / f z 1 / 3 , which stands for the heat transfer and friction characteristics. The best matching schemes of rotor-assembled strand, which significantly improves PEC to 2.01, are given by the regression analysis of uniform distribution and experimental design table. The single-factor experiments are performed to compare a tube installed with different kinds of rotor-assembled strands with a smooth tube without any strands when the Reynolds number changes between 20,000 and 60,000. The experimental result is in good agreement with the result obtained by the regression analysis of uniform distribution and experimental design. It is shown that the rotor diameters play important role in the heat transfer, and the optimal PEC value is obtained under the case that the rotor diameter is 21 mm. The rotor lead also contributes to the improvement of heat transfer and its optimal value is 700 mm in this study. The Nusselt number, friction factor and PEC increase with the increase in blade height. It shows that the uniform distribution and experimental design is an efficient method to find out the optimal parameters.

A Simple Heat of Crystallization Experiment.

Science.gov (United States)

De Nevers, Noel

1991-01-01

A demonstration used in a heat and material balances class that explains how a reusable heat pack works is described. An initial homework problem or exam question is provided with its solution. A discussion of the solution is included. (KR)

Reaction kinetics and reaction heat on thermal decomposition of solvent containing unstable reactive hydrocarbons with nitric acid at Tomsk-7 reprocessing plant

International Nuclear Information System (INIS)

Nishio, Gunji; Watanabe, Kouji; Koike, Tadao; Miyato, Teijiro.

1996-12-01

For analyzing a cause of the Tomsk-7 accident at Russian reprocessing plant, it is necessary to determine reaction-rate constant and reaction heat for a thermal decomposition of TBP/kerosine containing unstable reactive hydrocarbons with nitric acid. In JAERI, the rate constant and reaction heat were obtained from data measured with a differential thermal analyzer (DTA) for unstable hydrocarbons such as n-butanol, n-butyl nitrate, aromatic hydrocarbons, and cyclic compounds. The safety evaluation of Tomsk tank ruptured by the reaction was carried out by heat balance calculations between heat generation and heat loss in the tank using these rate constants and reaction heats. Consequently, it is clear that the cause of the tank rupture would be due to an exothermic reaction of aromatic hydrocarbons in kerosine made by petroleum with the concentrated nitric acid of 14.2N. (author)

Study on the CFD simulation of refrigerated container

Science.gov (United States)

Arif Budiyanto, Muhammad; Shinoda, Takeshi; Nasruddin

2017-10-01

The objective this study is to performed Computational Fluid Dynamic (CFD) simulation of refrigerated container in the container port. Refrigerated container is a thermal cargo container constructed from an insulation wall to carry kind of perishable goods. CFD simulation was carried out use cross sectional of container walls to predict surface temperatures of refrigerated container and to estimate its cooling load. The simulation model is based on the solution of the partial differential equations governing the fluid flow and heat transfer processes. The physical model of heat-transfer processes considered in this simulation are consist of solar radiation from the sun, heat conduction on the container walls, heat convection on the container surfaces and thermal radiation among the solid surfaces. The validation of simulation model was assessed uses surface temperatures at center points on each container walls obtained from the measurement experimentation in the previous study. The results shows the surface temperatures of simulation model has good agreement with the measurement data on all container walls.

The THETIS 80% blocked cluster experiment. Pt. 2

International Nuclear Information System (INIS)

Cooper, C.A.; Pearson, K.G.; Jowitt, D.

1984-09-01

Thermal-hydraulics experiments on a model PWR fuel assembly containing partial blockage are reported. The amount of heat removed by convection in single-phase flow is relatively small because a large fraction of the flow is diverted to bypass the blockage. Radial heat conduction across the blockage is an important heat transfer mechanism, particularly at low Reynolds numbers. Experimental results are in good agreement with predictions from the BERTHA code. (U.K.)

Foundations of ICRF heating--A historical perspective

International Nuclear Information System (INIS)

Hosea, J.C.

1994-01-01

Tom Stix has made many major contributions to the development of understanding of a wide array of rf heating and diagnostics methods, in experiment and theory. In recognition of his profound influence on ion cyclotron range of frequencies (ICRF) heating research, this paper is focused on two major building blocks contributed by him which served to help guide and quantify the research toward establishing ICRF heating as a viable technique for the reactor regime: (1) the formalism for quantitative evaluation of antenna loading contained in his 1962 text book and (2) his Fokker-Planck analysis for heating of ions and especially minority species ions in his 1975 Nuclear Fusion paper. Importantly, his work from the mid 1950s to the mid 1970s from which these two building blocks derive, provided a solid basis for the rapid developing ion cyclotron heating research in the 1970s and helped to guide that research to definitive demonstration of the viability of the minority ion heating regime as a reactor heating method by the end of the decade

Isotope distributions in primary heat transport and containment systems during a severe accident in CANDU type reactor

International Nuclear Information System (INIS)

Constantin, M.

2005-01-01

The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code. The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivation to begin with a simplified model. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU loss of coolant accident sequence (CATHENA code results). The source term of FPs introduced into the PHT was estimated by ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species were obtained by using SOPHAEROS module. The distributions within the containment are obtained by the CPA module (thermalhydraulic calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts of the containment and different hosts. (authors)

Facility with a nuclear district heating reactor

International Nuclear Information System (INIS)

Straub, H.

1988-01-01

The district heating reactor has a pressure vessel which contains the reactor core and at least one coolant conducting primary heat carrier surrounded by a heat sink. The pressure vessel has two walls with a space between them. This space is connected with a container which contains air as heat isolating medium and water as heat conducting medium. During the normal reactor operation the space is filled by air from the container with the aid of a blower, whereas in the case of a break-down of the cooling system it is filled by water which flows out of the container by gravity after the blower has been switched off. The after-heat, generated in the reactor core during cooling break-down, is removed into the heat sink surrounding the pressure vessel in a safe and simple way. 6 figs

The inaccuracy of heat transfer characteristics of insulated and non-insulated circular duct while neglecting the influence of heat radiation

International Nuclear Information System (INIS)

Hsien, T.-L.; Wong, K.-L.; Yu, S.-J.

2009-01-01

The non-insulated and insulated ducts are commonly applied in the industries and various buildings, because the heat radiation equation contains the 4th order exponential of temperature which is very complicate in calculations. Most heat transfer experts recognized from their own experiences that the heat radiation effect can be ignored due to the small temperature difference between insulated and non-insulated surface and surroundings. This paper studies in detail to check the inaccuracies of heat transfer characteristics non-insulated and insulated duct by comparing the results between considering and neglecting heat radiation effect. It is found that neglecting the heat radiation effect is likely to produce large errors of non-insulated and thin-insulated ducts in situations of ambient air with low external convection heat coefficients and larger surface emissivity, especially while the ambient air temperature is different from that of surroundings and greater internal fluid convection coefficients. It is also found in this paper that using greater duct surface emissivity can greatly improve the heat exchanger effect and using smaller insulated surface emissivity can obtain better insulation.

Review: heat pipe heat exchangers at IROST

OpenAIRE