The effect of a 48 h fast on the thermoregulatory responses to graded cooling in man.

Science.gov (United States)

Macdonald, I A; Bennett, T; Sainsbury, R

1984-10-01

The thermoregulatory responses to graded cooling were measured in 11 healthy male subjects after a 12 h fast and after a 48 h fast. The cooling stimulus was produced by changing the temperature of the skin of the trunk and legs with a water-perfused suit. Five levels of skin temperature from 35.5 to 24 degrees C were applied on each occasion. After a 12 h fast, core temperature was maintained during cooling. This maintenance of core temperature was associated with an increase in metabolic rate and a reduction in blood flow to the hand and to the forearm. After 48 h of fasting, the subjects could not maintain core temperature during cooling, and a decrease of 0.36 +/- 0.05 degrees C occurred as the suit temperature was reduced from 35.9 to 24 degrees C. Metabolic rate was slightly higher after the 48 h fast than after the 12 h fast, but similar increases in metabolic rate were observed during cooling. Vasoconstriction in the hand was initially less after a 48 h fast than after a 12 h fast, but at the lowest suit temperature, hand blood flow was similar, and low, on both occasions. After 48 h of fasting, forearm blood flow was elevated at all suit temperatures, being approximately twice the level recorded after the 12 h fast. Venous plasma noradrenaline levels did not change during cooling after the 12 h fast, whilst after 48 h of fasting a significant increase in noradrenaline level was observed at the lowest suit temperature. The results of this study provide further evidence that fasting induces an impairment of autonomic reflex mechanisms, but it is not clear whether this is due to a suppression of sympathetic nervous activity.

Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

International Nuclear Information System (INIS)

Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

2001-01-01

Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

International Nuclear Information System (INIS)

Sackett, J.I.

1983-01-01

This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

Water vapor as a perspective coolant for fast reactors

International Nuclear Information System (INIS)

Kalafati, D.D.; Petrov, S.I.

1978-01-01

Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

Study on neutronics performance of flower shape advanced supercritical water cooled fast reactor with different solid moderators

International Nuclear Information System (INIS)

Yu Tao; Li Zhifeng; Xie Jinsen; Peng Honghua

2015-01-01

The supercritical water cooled fast reactors worked at such harsh condition with high temperature and high pressure, huge hydrogen balance pressure and thermal shock can result in a great loss of hydrogen. The released hydrogen would be out of control under accident situations. K_e_f_f, conversion ratio, moderator temperature effect, Doppler effect and void effect of different material such as ZrH_1_._7, Bp, BeO, C and SiC are discussed. BeO and SiC hold better integrated performance among these materials. Besides, moderators have less effect on the Doppler effect of fuel. (authors)

Thermal analysis experiment for elucidating sodium-water chemical reaction mechanism in steam generator of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki

2012-01-01

For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature, and decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of sodium monoxide (Na 2 O) generation was obtained. Thermal analysis results indicated that Na 2 O generation at the secondary overall reaction should be considered during the sodium-water reaction. (author)

Analysis on small long life reactor using thorium fuel for water cooled and metal cooled reactor types

International Nuclear Information System (INIS)

Permana, Sidik

2009-01-01

Long-life reactor operation can be adopted for some special purposes which have been proposed by IAEA as the small and medium reactor (SMR) program. Thermal reactor and fast reactor types can be used for SMR and in addition to that program the utilization of thorium fuel as one of the candidate as a 'partner' fuel with uranium fuel which can be considered for optimizing the nuclear fuel utilization as well as recycling spent fuel. Fissile U-233 as the main fissile material for thorium fuel shows higher eta-value for wider energy range compared with other fissile materials of U-235 and Pu-239. However, it less than Pu-239 for fast energy region, but it still shows high eta-value. This eta-value gives the reactor has higher capability for obtaining breeding condition or high conversion capability. In the present study, the comparative analysis on small long life reactor fueled by thorium for different reactor types (water cooled and metal cooled reactor types). Light water and heavy water have been used as representative of water-cooled reactor types, and for liquid metal-cooled reactor types, sodium-cooled and lead-bismuth-cooled have been adopted. Core blanket arrangement as general design configuration, has been adopted which consist of inner blanket region fueled by thorium oxide, and two core regions (inner and out regions) fueled by fissile U-233 and thorium oxide with different percentages of fissile content. SRAC-CITATION and JENDL-33 have been used as core optimization analysis and nuclear data library for this analysis. Reactor operation time can reaches more than 10 years operation without refueling and shuffling for different reactor types and several power outputs. As can be expected, liquid metal cooled reactor types can be used more effective for obtaining long life reactor with higher burnup, higher power density, higher breeding capability and lower excess reactivity compared with water-cooled reactors. Water cooled obtains long life core operation

Lead- or Lead-bismuth-cooled fast reactors

International Nuclear Information System (INIS)

Bouchter, J.C.; Courouau, J.L.; Dufour, P.; Guidez, J.; Latge, C.; Martinelli, L.; Renault, C.; Rimpault, G.

2014-01-01

Lead-cooled fast reactors are one of the 6 concepts retained for the 4. generation of nuclear reactors. So far no lead-cooled reactors have existed in the world except lead-bismuth-cooled reactors in soviet submarines. Some problems linked to the use of the lead-bismuth eutectic appeared but were satisfactorily solved by a more rigorous monitoring of the chemistry of the lead-bismuth coolant. Lead presents various advantages as a coolant: no reactivity with water and the air,a high boiling temperature and low contamination when irradiated. The main asset of the lead-bismuth alloy is the drop of the fusion temperature from 327 C degrees to 125 C degrees. The main drawback of using lead (or lead-bismuth) is its high corrosiveness with metals like iron, chromium and nickel. The high corrosiveness of the coolant implies low flow velocities which means a bigger core and consequently a bigger reactor containment. Different research programs in the world (in Europe, Russia and the USA) are reviewed in the article but it appears that the development of this type of reactor requires technological breakthroughs concerning materials and the resistance to corrosion. Furthermore the concept of lead-cooled reactors seems to be associated to a range of low output power because of the compromise between the size of the reactor and its resistance to earthquakes. (A.C.)

Water cooling system for sintering furnaces of nuclear fuel pellets

International Nuclear Information System (INIS)

1996-01-01

This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author)

Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

International Nuclear Information System (INIS)

Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

2009-01-01

This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

Cooled Water Production System,

Science.gov (United States)

The invention refers to the field of air conditioning and regards an apparatus for obtaining cooled water . The purpose of the invention is to develop...such a system for obtaining cooled water which would permit the maximum use of the cooling effect of the water -cooling tower.

Air-cooled fast discharge resistors for ITER magnets

International Nuclear Information System (INIS)

Tanchuk, Victor; Grigoriev, Sergey; Lokiev, Vladimir; Roshal, Alexander; Song, Inho; Buzykin, Oleg

2011-01-01

The ITER superconducting magnets will store up to 50 GJ of magnetic energy per operation cycle. In case of coil quench the energy stored in the coils must be extracted rapidly with a time constant from 7.5 to 14 s. It will be achieved by fast discharge resistors (FDR) normally bridged by circuit breakers and inserted in series with the superconducting coils. The fast discharge of the coils results practically in adiabatic heating of the resistive elements up to 200-300 deg. C. The resistors need to be cooled to the initial temperature over 6-8 h. Natural air circulation is proposed as a cooling method. In order to simulate the temperature response of the resistors to energy released in the resistive plates and to demonstrate their cooling capability within the required time by natural air circulation the numerical model of the resistor cooling circuit has been developed. As the calculations have shown, the developed FDR cooling system based on cooling by natural air circulation is capable of providing the required temperature operation regime of FDRs, but the supply channels are to be optimized so that the cooling time does not exceed the permissible one.

Delayed gamma power measurement for sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

2011-01-15

Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

Containment atmosphere cooling system for experimental fast reactor 'JOYO'

International Nuclear Information System (INIS)

Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

1979-01-01

The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

On the substantion of permissible concentrations of plutonium isotopes in the water of fresh water and sea water NPP cooling reservoirs

International Nuclear Information System (INIS)

Grachev, M.I.; Gusev, D.I.; Stepanova, V.D.

1985-01-01

Substantiation of maximum permissible concentration (PC) of plutonium isotopes ( 238 Pu, 239 Pu, 240 Pu) in fresh and sea water cooling reservoirs of NPP with fast neutron reactors is given. The main criterion when calculating permissible plutonium content in water of surface reservoirs is the requirement not to exceed the established limits for radiation doses to persons resulted from water use. Data on coefficients of plutonium concentration in sea and fresh water hydrobionts are presented as well as on plutonium PC in water of fresh and sea water cooling reservoirs and bottom sediments of sea water cooling reservoirs. It is shown that doses to critical groups of population doesn't exceed potentially hazardous levels due to plutonium intake through food chains. But the calculation being carried out further should be corrected

Cooling water distribution system

Science.gov (United States)

Orr, Richard

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

International Nuclear Information System (INIS)

2012-12-01

The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

Water cooling coil

Energy Technology Data Exchange (ETDEWEB)

Suzuki, S; Ito, Y; Kazawa, Y

1975-02-05

Object: To provide a water cooling coil in a toroidal nuclear fusion device, in which coil is formed into a small-size in section so as not to increase dimensions, weight or the like of machineries including the coil. Structure: A conductor arranged as an outermost layer of a multiple-wind water cooling coil comprises a hollow conductor, which is directly cooled by fluid, and as a consequence, a solid conductor disposed interiorly thereof is cooled indirectly.

Study on the Effect of water Injection Momentum on the Cooling Effect of Rocket Engine Exhaust Plume

Science.gov (United States)

Yang, Kan; Qiang, Yanhui; Zhong, Chenghang; Yu, Shaozhen

2017-10-01

For the study of water injection momentum factors impact on flow field of the rocket engine tail flame, the numerical computation model of gas-liquid two phase flow in the coupling of high temperature and high speed gas flow and low temperature liquid water is established. The accuracy and reliability of the numerical model are verified by experiments. Based on the numerical model, the relationship between the flow rate and the cooling effect is analyzed by changing the water injection momentum of the water spray pipes. And the effective mathematical expression is obtained. What’s more, by changing the number of the water spray and using small flow water injection, the cooling effect is analyzed to check the application range of the mathematical expressions. The results show that: the impact and erosion of the gas flow field could be reduced greatly by water injection, and there are two parts in the gas flow field, which are the slow cooling area and the fast cooling area. In the fast cooling area, the influence of the water flow momentum and nozzle quantity on the cooling effect can be expressed by mathematical functions without causing bifurcation flow for the mainstream gas. The conclusion provides a theoretical reference for the engineering application.

Analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors

International Nuclear Information System (INIS)

Schultheiss, G.F.

1976-09-01

The aim of this analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors is to get results on the possible extent of blockages and the time necessary for growth which may be used for a safety evaluation. After an introduction where the thermohydraulic and physical/chemical aspects of the problems are considered, the causes for the local cooling disturbances and the phenomena arising with it are freated in more detail. (orig./TK) [de

Fuel Development For Gas-Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

M. K. Meyer

2006-06-01

The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

Cooling clothing utilizing water evaporation

DEFF Research Database (Denmark)

Sakoi, Tomonori; Tominaga, Naoto; Melikov, Arsen Krikor

2014-01-01

. To prevent wet discomfort, the T-shirt was made of a polyester material having a water-repellent silicon coating on the inner surface. The chest, front upper arms, and nape of the neck were adopted as the cooling areas of the human body. We conducted human subject experiments in an office with air......We developed cooling clothing that utilizes water evaporation to cool the human body and has a mechanism to control the cooling intensity. Clean water was supplied to the outer surface of the T-shirt of the cooling clothing, and a small fan was used to enhance evaporation on this outer surface...... temperature ranging from 27.4 to 30.7 °C to establish a suitable water supply control method. A water supply control method that prevents water accumulation in the T-shirt and water dribbling was validated; this method is established based on the concept of the water evaporation capacity under the applied...

Materials challenges for the supercritical water-cooled reactor (SCWR)

International Nuclear Information System (INIS)

Baindur, S.

2008-01-01

This paper discusses the materials requirements of the Supercritical Water-cooled Reactor (SCWR) which arise from its severe expected operating conditions: (i) Outlet Temperature (to 650 C); (ii) Pressure of 25 MPa for the coolant containment, (iii) Thermochemical stress in the presence of supercritical water, and (iv) Radiative damage (up to 150 dpa for the fast spectrum variant). These operating conditions are reviewed; the phenomenology of materials in the supercritical water environment that create the materials challenges is discussed; knowledge gaps are identified, and efforts to understand material behaviour under the operating conditions expected in the SCWR are described. (author)

Computational methodology of sodium-water reaction phenomenon in steam generator of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Takata, Takashi; Yamaguchi, Akira; Uchibori, Akihiro; Ohshima, Hiroyuki

2009-01-01

A new computational methodology of sodium-water reaction (SWR), which occurs in a steam generator of a liquid-sodium-cooled fast reactor when a heat transfer tube in the steam generator fails, has been developed considering multidimensional and multiphysics thermal hydraulics. Two kinds of reaction models are proposed in accordance with a phase of sodium as a reactant. One is the surface reaction model in which water vapor reacts directly with liquid sodium at the interface between the liquid sodium and the water vapor. The reaction heat will lead to a vigorous evaporation of liquid sodium, resulting in a reaction of gas-phase sodium. This is designated as the gas-phase reaction model. These two models are coupled with a multidimensional, multicomponent gas, and multiphase thermal hydraulics simulation method with compressibility (named the 'SERAPHIM' code). Using the present methodology, a numerical investigation of the SWR under a pin-bundle configuration (a benchmark analysis of the SWAT-1R experiment) has been carried out. As a result, the maximum gas temperature of approximately 1,300degC is predicted stably, which lies within the range of previous experimental observations. It is also demonstrated that the maximum temperature of the mass weighted average in the analysis agrees reasonably well with the experimental result measured by thermocouples. The present methodology will be promising to establish a theoretical and mechanical modeling of secondary failure propagation of heat transfer tubes due to such as an overheating rupture and a wastage. (author)

Passive containment cooling water distribution device

Science.gov (United States)

Conway, Lawrence E.; Fanto, Susan V.

1994-01-01

A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using a series of radial guide elements and cascading weir boxes to collect and then distribute the cooling water into a series of distribution areas through a plurality of cascading weirs. The cooling water is then uniformly distributed over the curved surface by a plurality of weir notches in the face plate of the weir box.

Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

Directory of Open Access Journals (Sweden)

Sungjoo Lee

2016-09-01

Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

Gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Yoshida, Hiroyuki

1982-07-01

Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

Cooling-water amounts, temperature, and the environment

International Nuclear Information System (INIS)

Koops, F.B.J.; Donze, M.; Hadderingh, R.H.

1979-01-01

The release of heat from power plants into a water can take place with relative small quantities of cooling water, highly warmed up accordingly, or with large quantities of cooling water slightly warmed up. The utilization of cooling water is bound to certain guidelines established by the authorities. With the intention to protect the environment, the admissable temperatures and warming-up have been strictly limited by the authorities. In the Netherlands, we have presently temporary cooling water guidelines which allow a max. temperature of the cooling water in the cooling cycle of 30 0 C and a maximum admissible temperature rise in the condenser between 7 0 C during summer and 15 0 C during winter. It has also been determined in these requirements how much cooling water at least has to be used to discharge a specified quantity of heat. Plankton, spawn and young fish are dragged with the cooling water. Harm to these organisms can be caused mechanically by pumps, sieves and the condenser or they can be harmed by the temperature rise in the condenser. Investigations showed that mechanical harm to spawn and young fish in the cooling water flow should not be ignored, and that detectable harm to plankton organisms takes place only at water temperatures above 32 0 C. The cooling water consumption can therefore be optimised as follows: The solution of a greater temperature increase and a slightly higher value for the temperature maximum can reduce the cooling water quantity. This reduction of the cooling water quantity reduces the destruction of the fish quantity, which gets into the cooling water system, especially during the summer. If the temperature rise and the temperature itself are not selected too high, the destruction of fish may be reduced without causing serious damage to the plankton. (orig.) [de

Emergency cooling method and system for gas-cooled nuclear reactors

International Nuclear Information System (INIS)

Kumpf, H.

1982-01-01

For emergency cooling of gas-cooled fast breeder reactors (GSB), which have a core consisting of a fission zone and a breeding zone, water is sprayed out of nozzles on to the core from above in the case of an incident. The water which is not treated with boron is taken out of a reservoir in the form of a storage tank in such a maximum quantity that the cooling water gathering in the space below the core rises at most up to the lower edge of the fission zone. (orig./GL) [de

Cooling water systems design using process integration

CSIR Research Space (South Africa)

Gololo, KV

2010-09-01

Full Text Available Cooling water systems are generally designed with a set of heat exchangers arranged in parallel. This arrangement results in higher cooling water flowrate and low cooling water return temperature thus reducing cooling tower efficiency. Previous...

Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

NARCIS (Netherlands)

van Rooijen, W.F.G.

2006-01-01

This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

Mycobacteria in Finnish cooling tower waters.

Science.gov (United States)

Torvinen, Eila; Suomalainen, Sini; Paulin, Lars; Kusnetsov, Jaana

2014-04-01

Evaporative cooling towers are water systems used in, e.g., industry and telecommunication to remove excess heat by evaporation of water. Temperatures of cooling waters are usually optimal for mesophilic microbial growth and cooling towers may liberate massive amounts of bacterial aerosols. Outbreaks of legionellosis associated with cooling towers have been known since the 1980's, but occurrences of other potentially pathogenic bacteria in cooling waters are mostly unknown. We examined the occurrence of mycobacteria, which are common bacteria in different water systems and may cause pulmonary and other soft tissue infections, in cooling waters containing different numbers of legionellae. Mycobacteria were isolated from all twelve cooling systems and from 92% of the 24 samples studied. Their numbers in the positive samples varied from 10 to 7.3 × 10(4) cfu/L. The isolated species included M. chelonae/abscessus, M. fortuitum, M. mucogenicum, M. peregrinum, M. intracellulare, M. lentiflavum, M. avium/nebraskense/scrofulaceum and many non-pathogenic species. The numbers of mycobacteria correlated negatively with the numbers of legionellae and the concentration of copper. The results show that cooling towers are suitable environments for potentially pathogenic mycobacteria. Further transmission of mycobacteria from the towers to the environment needs examination. © 2013 APMIS. Published by John Wiley & Sons Ltd.

Demineralised water cooling in the LHC accelerator

CERN Document Server

Peón-Hernández, G

2002-01-01

In spite of the LHC accelerator being a cryogenic machine, it remains nevertheless a not negligible heat load to be removed by conventional water-cooling. About 24MW will be taken away by demineralised water cooled directly by primary water from the LHC cooling towers placed at the even points. This paper describes the demineralised water network in the LHC tunnel including pipe diameters, lengths, water speed, estimated friction factor, head losses and available supply and return pressures for each point. It lists all water cooled equipment, highlights the water cooled cables as the most demanding equipment followed by the radio frequency racks and cavities, and by the power converters. Their main cooling requirements and their positions in the tunnel are also presented.

Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

International Nuclear Information System (INIS)

Kotake, Syoji; Nishikawa, Akira

2005-01-01

Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

Once-through cycle, supercritical-pressure light water cooled reactor concept

Energy Technology Data Exchange (ETDEWEB)

Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

2001-07-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through cycle, supercritical-pressure light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.

2001-01-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

18 CFR 420.44 - Cooling water.

Science.gov (United States)

2010-04-01

... 18 Conservation of Power and Water Resources 2 2010-04-01 2010-04-01 false Cooling water. 420.44 Section 420.44 Conservation of Power and Water Resources DELAWARE RIVER BASIN COMMISSION ADMINISTRATIVE MANUAL BASIN REGULATIONS-WATER SUPPLY CHARGES Charges; Exemptions § 420.44 Cooling water. Water used...

SSTAR: The US lead-cooled fast reactor (LFR)

International Nuclear Information System (INIS)

Smith, Craig F.; Halsey, William G.; Brown, Neil W.; Sienicki, James J.; Moisseytsev, Anton; Wade, David C.

2008-01-01

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system

Gas cooled fast reactor research and development program

International Nuclear Information System (INIS)

Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

1980-03-01

The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

Gas Cooled Fast Reactors: Recent advances and prospects

International Nuclear Information System (INIS)

Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

2013-01-01

Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

Status of liquid metal cooled fast reactor technology

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-04-01

During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants Refs, figs, tabs

Status of liquid metal cooled fast reactor technology

International Nuclear Information System (INIS)

1999-04-01

During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

Reducing water consumption of an industrial plant cooling unit using hybrid cooling tower

International Nuclear Information System (INIS)

Rezaei, Ebrahim; Shafiei, Sirous; Abdollahnezhad, Aydin

2010-01-01

Water consumption is an important problem in dry zones and poor water supply areas. For these areas use of a combination of wet and dry cooling towers (hybrid cooling) has been suggested in order to reduce water consumption. In this work, wet and dry sections of a hybrid cooling tower for the estimation of water loss was modeled. A computer code was also written to simulate such hybrid cooling tower. To test the result of this simulation, a pilot hybrid tower containing a wet tower and 12 compact air cooled heat exchangers was designed and constructed. Pilot data were compared with simulation data and a correction factor was added to the simulation. Ensuring that the simulation represents the actual data, it was applied to a real industrial case and the effect of using a dry tower on water loss reduction of this plant cooling unit was investigated. Finally feasibility study was carried out to choose the best operating conditions for the hybrid cooling tower configuration proposed for this cooling unit.

Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

International Nuclear Information System (INIS)

Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

2016-09-01

The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

Desalting a process cooling water using nanofiltration

NARCIS (Netherlands)

Radier, R.G.J.; van Oers, C.W.; Steenbergen, A.; Wessling, Matthias

2001-01-01

The cooling water system of a chemical plant of Akzo Nobel is a partly open system. The site is located at the North Sea. The air in contact with the cooling water contains seawater droplets dissolving and increasing the chloride concentration. The cooling water contains chromate to protect the

Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-03-01

This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Gatley, J.A.

1979-01-01

Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Thatcher, G.; Mitchell, A.J.

1981-01-01

Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

International Nuclear Information System (INIS)

Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

2001-01-01

The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

Core Seismic Tests for a Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Koo, Gyeong Hoi; Lee, J. H

2007-01-15

This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

Assessment of gas cooled fast reactor with indirect supercritical CO2 cycle

International Nuclear Information System (INIS)

Hejzlar, P.; Driscoll, M. J.; Dostal, V.; Dumaz, P.; Poullennec, G.; Alpy, N.

2006-01-01

Various indirect power cycle options for a helium cooled Gas cooled Fast Reactor (GFR) with particular focus on a supercritical CO 2 (SCO 2 ) indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The Balance Of Plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and SCO 2 recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of 550 .deg. C, (2) advanced design with turbine inlet temperature of 650 .deg. C and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect SCO 2 recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR 'proximate-containment' and the BOP for the SCO 2 cycle is very compact. Both these factors will lead to reduced capital cost

Safety Design Criteria of Indian Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

2013-01-01

• Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

Materials science research for sodium cooled fast reactors

Indian Academy of Sciences (India)

The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

Emergency cooling of presurized water reactor

International Nuclear Information System (INIS)

Sykora, D.

1981-01-01

The method described of emergency core cooling in the pressurized water reactor is characterized by the fact that water is transported to the disturbed primary circuit or direct to the reactor by the action of the energy and mass of the steam and/or liquid phase of the secondary circuit coolant, which during emergency core cooling becomes an emergency cooling medium. (B.S.)

Cooling water requirements and nuclear power plants

International Nuclear Information System (INIS)

Rao, T.S.

2010-01-01

Indian nuclear power programme is poised to scuttle the energy crisis of our time by proposing joint ventures for large power plants. Large fossil/nuclear power plants (NPPs) rely upon water for cooling and are therefore located near coastal areas. The amount of water a power station uses and consumes depends on the cooling technology used. Depending on the cooling technology utilized, per megawatt existing NPPs use and consume more water (by a factor of 1.25) than power stations using other fuel sources. In this context the distinction between 'use' and 'consume' of water is important. All power stations do consume some of the water they use; this is generally lost as evaporation. Cooling systems are basically of two types; Closed cycle and Once-through, of the two systems, the closed cycle uses about 2-3% of the water volumes used by the once-through system. Generally, water used for power plant cooling is chemically altered for purposes of extending the useful life of equipment and to ensure efficient operation. The used chemicals effluent will be added to the cooling water discharge. Thus water quality impacts on power plants vary significantly, from one electricity generating technology to another. In light of massive expansion of nuclear power programme there is a need to develop new ecofriendly cooling water technologies. Seawater cooling towers (SCT) could be a viable option for power plants. SCTs can be utilized with the proper selection of materials, coatings and can achieve long service life. Among the concerns raised about the development of a nuclear power industry, the amount of water consumed by nuclear power plants compared with other power stations is of relevance in light of the warming surface seawater temperatures. A 1000 MW power plant uses per day ∼800 ML/MW in once through cooling system; while SCT use 27 ML/MW. With the advent of new marine materials and concrete compositions SCT can be constructed for efficient operation. However, the

Neutronic design for a 100MWth Small modular natural circulation lead or lead-alloy cooled fast reactors core

International Nuclear Information System (INIS)

Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q.

2015-01-01

Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW th natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

Device for recirculation cooling of cooling water by natural or forced chaft

Energy Technology Data Exchange (ETDEWEB)

Ruehl, H; Honekamp, H; Katzmann, A

1975-10-23

The invention is concerned with a device for recirculation cooling of cooling water by natural or forced draft. Through a cascading system mounted on supporting columns at a vertical distance to ground level, cooling air is flowing in cross- or counterflow to the cooling water freely falling from the cascading system. The cooling water collecting zone below the cascading system has an absorption floor arranged nearly horizontal and/or inclined, with a cam-type profile on its upperside, which is bounded on its circumference by at least one cooling water release channel provided below its level and/or which is divided in the sense of a surface subdivision. By these means, a reduction of the amount of material required for the supporting columns and an increase of the stability of the columns is to be achieved. Furthermore, the deposition of mud is to be avoided as for as possible, and noise generation during operation is to be reduced considerably. For this purpose, the absorption floor may be made of material sound insulating and/or may be coated with such a material.

Open air-vapor compression refrigeration system for air conditioning and hot water cooled by cool water

International Nuclear Information System (INIS)

Hou Shaobo; Li Huacong; Zhang Hefei

2007-01-01

This paper presents an open air-vapor compression refrigeration system for air conditioning and hot water cooled by cool water and proves its feasibility through performance simulation. Pinch technology is used in analysis of heat exchange in the surface heat exchanger, and the temperature difference at the pinch point is selected as 6 o C. Its refrigeration depends mainly on both air and vapor, more efficient than a conventional air cycle, and the use of turbo-machinery makes this possible. This system could use the cool in the cool water, which could not be used to cool air directly. Also, the heat rejected from this system could be used to heat cool water to 33-40 o C. The sensitivity analysis of COP to η c and η t and the simulated results T 4 , T 7 , T 8 , q 1 , q 2 and W m of the cycle are given. The simulations show that the COP of this system depends mainly on T 7 , η c and η t and varies with T 3 or T wet and that this cycle is feasible in some regions, although the COP is sensitive to the efficiencies of the axial compressor and turbine. The optimum pressure ratio in this system could be lower, and this results in a fewer number of stages of the axial compressor. Adjusting the rotation speed of the axial compressor can easily control the pressure ratio, mass flow rate and the refrigerating capacity. The adoption of this cycle will make the air conditioned room more comfortable and reduce the initial investment cost because of the obtained very low temperature air. Humid air is a perfect working fluid for central air conditioning and no cost to the user. The system is more efficient because of using cool water to cool the air before the turbine. In addition, pinch technology is a good method to analyze the wet air heat exchange with water

Johnson screen for cooling water intakes

International Nuclear Information System (INIS)

Cook, L.E.

1978-01-01

Johnson surface-water screens provide an alternative to vertical traveling screens for power plant cooling water intakes. In this paper, flow field modeling is discussed, and a series of case studies is presented. The hydraulic information obtained is discussed as it applies to the exclusion of biota and debris from cooling water intake systems

Storage of HLW in engineered structures: air-cooled and water-cooled concepts

International Nuclear Information System (INIS)

Ahner, S.; Dekais, J.J.; Puttke, B.; Staner, P.

1981-01-01

A comparative study on an air-cooled and a water-cooled intermediate storage of vitrified, highly radioactive waste (HLW) in overground installations has been performed by Nukem and Belgonucleaire respectively. In the air-cooled storage concept the decay heat from the storage area will be removed using natural convection. In the water-cooled storage concept the decay heat is carried off by a primary and secondary forced-cooling system with redundant and diverse devices. The safety study carried out by Nukem used a fault tree method. It shows that the reliability of the designed water-cooled system is very high and comparable to the inherent, safe, air-cooled system. The impact for both concepts on the environment is determined by the release route, but even during accident conditions the release is far below permissible limits. The economic analysis carried out by Belgonucleaire shows that the construction costs for both systems do not differ very much, but the operation and maintenance costs for the water-cooled facility are higher than for the air cooled facility. The result of the safety and economic analysis and the discussions with the members of the working group have shown some possible significant modifications for both systems, which are included in this report. The whole study has been carried out using certain national criteria which, in certain Member States at least, would lead to a higher standard of safety than can be justified on any social, political or economic grounds

Environmental and legal aspects of cooling water chemistry

International Nuclear Information System (INIS)

Hoffmann, H.J.

1988-01-01

The discharge and management of cooling water and waste water are subject to a number of ecological and legal requirements. For example, waste heat and cooling water constituents may affect surface bodies of water, or waste water discharge may have adverse effects on surface water and ground water. Waste water and cooling water discharge are subject to the Water Management Act (WHG) and the Waste Water Act, with about 50 administrative regulations. The requirements on water chemistry and analysis are gone into. (orig./HP) [de

Design considerations for economically competitive sodium cooled fast reactors

International Nuclear Information System (INIS)

Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

2009-01-01

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

Under-Sodium-Viewing as one technique for periodic inspections in sodium-cooled fast reactors-- possibilities and limits

International Nuclear Information System (INIS)

Weiss, H.

1979-07-01

Periodic inspections are gaining increasingly technical importance for fast sodium cooled reactors. Among others the reactor tank and its internals have to be inspected, whereby licensing experts partly are requesting the standards of Light Water Reactors. This leads to difficulties in sodium cooled reactors because of the non-transparent coolant sodium and their compact structure. In order to avoid the complete dumping of the sodium, the under sodium viewing shall be applied besides other inspection methods. Since this is a new method, which is still in its development phase, this report presents and discusses the technical and physical basis and outlines possibilities and limits [de

Selection of steam generator materials for sodium cooled fast breeders

International Nuclear Information System (INIS)

Berge, P.

1977-01-01

The sodium water heat exchangers are now considered as the stumbling block in the development of liquid metal cooled fast breeders, due to the risk of sodium-water reactions. The selection of the materials for these tube-bundles has been very broad, for the different existing, or in-project, reactors in the world: low alloy 2 1/4 Cr - 1 Mo steels (unstabilized or stabilized); 9 Cr - 1 Mo ferritic steel; 18 Cr - 10 Ni austenitic stainless steels; alloy 800. On can also add other ferritic steels, as 9 Cr - 2 Mo stabilized, which are studied for this application. In the framework of the E.D.F.-C.E.A. working group a major effort was undertaken to study the characteristics of these various materials with respect to the main criteria governing construction of the tube bundles and their performance in service: mechanical characteristics at high temperature; fabrication and welding; behavior with respect to mass transfer in sodium; carburization and decarburization; corrosion resistance. The main lines and results of this program are described [fr

Sea water take-up facility for cooling reactor auxiliary

International Nuclear Information System (INIS)

Numata, Noriko; Mizutani, Akira; Hirako, Shizuka; Uchiyama, Yuichi; Oda, Atsushi.

1997-01-01

The present invention provides an improvement of a cooling sea water take-up facility for cooling auxiliary equipments of nuclear power plant. Namely, an existent sea water take-up facility for cooling reactor auxiliary equipments has at least two circulation water systems and three independent sea water systems for cooling reactor auxiliary equipments. In this case, a communication water channel is disposed, which connects the three independent sea water systems for cooling reactor auxiliary equipments mutually by an opening/closing operation of a flow channel partitioning device. With such a constitution, even when any combination of two systems among the three circulation water systems is in inspection at the same time, one system for cooling the reactor auxiliary equipments can be kept operated, and one system is kept in a stand-by state by the communication water channel upon periodical inspection of water take-up facility for cooling the auxiliary equipments. As a result, the sea water take-up facility for cooling auxiliary equipments of the present invention have operation efficiency higher than that of a conventional case while keeping the function and safety at the same level as in the conventional case. (I.S.)

Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

International Nuclear Information System (INIS)

Ninokata, H.; Kamide, H.

2011-01-01

In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

Shape optimization of a sodium cooled fast reactor

International Nuclear Information System (INIS)

Schmitt, D.; Allaire, G.; Pantz, O.; Pozin, N.

2013-01-01

Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth. Usual optimization methods for core conception are based on a parametric description of a given core design. New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints. First studies show that these methods could be applied to sodium cooled core conception. In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a given realistic core layout. Its characteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas. (authors)

Legionella confirmation in cooling tower water

Science.gov (United States)

Farhat, Maha; Shaheed, Raja A.; Al-Ali, Haidar H.; Al-Ghamdi, Abdullah S.; Al-Hamaqi, Ghadeer M.; Maan, Hawraa S.; Al-Mahfoodh, Zainab A.; Al-Seba, Hussain Z.

2018-01-01

Objectives: To investigate the presence of Legionella spp in cooling tower water. Legionella proliferation in cooling tower water has serious public health implications as it can be transmitted to humans via aerosols and cause Legionnaires’ disease. Methods: Samples of cooling tower water were collected from King Fahd Hospital of the University (KFHU) (Imam Abdulrahman Bin Faisal University, 2015/2016). The water samples were analyzed by a standard Legionella culture method, real-time polymerase chain reaction (RT-PCR), and 16S rRNA next-generation sequencing. In addition, the bacterial community composition was evaluated. Results: All samples were negative by conventional Legionella culture. In contrast, all water samples yielded positive results by real-time PCR (105 to 106 GU/L). The results of 16S rRNA next generation sequencing showed high similarity and reproducibility among the water samples. The majority of sequences were Alpha-, Beta-, and Gamma-proteobacteria, and Legionella was the predominant genus. The hydrogen-oxidizing gram-negative bacterium Hydrogenophaga was present at high abundance, indicating high metabolic activity. Sphingopyxis, which is known for its resistance to antimicrobials and as a pioneer in biofilm formation, was also detected. Conclusion: Our findings indicate that monitoring of Legionella in cooling tower water would be enhanced by use of both conventional culturing and molecular methods. PMID:29436561

Cooling tower water ozonation at Southern University

International Nuclear Information System (INIS)

Chen, C.C.; Knecht, A.T.; Trahan, D.B.; Yaghi, H.M.; Jackson, G.H.; Coppenger, G.D.

1990-01-01

Cooling-tower water is a critical utility for many industries. In the past, inexpensive water coupled with moderate regulation of discharge water led to the neglect of the cooling tower as an energy resource. Now, with the increased cost of chemical treatment and tough EPA rules and regulations, this situation is rapidly changing. The operator of the DOE Y-12 Plant in Oak Ridge as well as many other industries are forced to develop an alternate method of water treatment. The cooling tower is one of the major elements in large energy systems. The savings accrued from a well engineered cooling tower can be a significant part of the overall energy conservation plan. During a short-term ozonation study between 1987-1988, the Y-12 Plant has been successful in eliminating the need for cooling tower treatment chemicals. However, the long-term impact was not available. Since April 1988, the ozone cooling water treatment study at the Y-12 Plant has been moved to the site at Southern University in Baton Rouge, Louisiana. The purpose of this continued study is to determine whether the use of ozonation on cooling towers is practical from an economic, technical and environmental standpoint. This paper discusses system design, operating parameter and performance testing of the ozonation system at Southern University

Materials Options of Steam Generator for Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Fu Xiaogang; Long Bin; Han Liqing; Qin Bo; Zhang Jinquan; Wang Shuxing

2013-01-01

Overview of the material options of steam generator for sodium-cooled fast reactors, the method to calculate the service life, the thinning of wall thickness and the sodium corrosion rate, the degradation of mechanical properties (thermal aging and decarburization) and the calculation results of theoretical models

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

International Nuclear Information System (INIS)

Muramatu, Toshiharu; Yamaguchi, Akira

2004-01-01

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

Development Status on Innovative Sodium-Cooled Fast Reactor (JSFR)

International Nuclear Information System (INIS)

Yanagisawa, Tsutomu; Sato, Kazujiro

2006-01-01

The first step in Japan's nuclear fuel cycle policy is to introduce MOX recycle in light water reactors (LWRs) and the final step is to establish multiple TRU recycle in fast reactors (FRs), with the goal of realizing a stable supply, effective use of nuclear fuel resources, and the environmentally friendly production of energy. Therefore, a feasibility study on commercialized FR cycle systems has been launched since July 1999 by a Japanese joint project team of Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC: the representative of the electric utilities) in cooperation with Central Research Institute of Electric Power Industry (CRIEPI) and vendors. In the period from July 1999 to March 2001, the feasibility study phase-I was conducted to screen out representative FR cycle concepts. In the feasibility study phase-II (April 2001 - March 2006), investigations in to the representative FR concepts were carried out to clarify the most promising concept for commercial deployment. This paper describes an innovative sodium-cooled FR, which is named as the JAEA Sodium-cooled FR (JSFR), as the most promising FR concept that meets the Generation-IV performance target. The JSFR employs several advanced technologies, such as an oxide dispersion strengthened (ODS) cladding for higher burn-up, a short-piping configuration with less elbows by adopting high chromium steel, a large scale integrated intermediate heat exchanger with a primary circulation pump, etc. Based on the design, construction and operation experiences of JOYO and MONJU, there are extensive technology bases for sodium-cooled FRs. Nevertheless, several innovative technologies implemented into the JSFR have to be developed in order to realize higher economic competitiveness by reducing construction costs and improving plant availability

Cooling water conditioning and quality control for tokamaks

International Nuclear Information System (INIS)

Gootgeld, A.M.

1995-01-01

Designers and operators of Tokamaks and all associated water cooled, peripheral equipment, are faced with the task of providing and maintaining closed-loop, low conductivity, low impurity, cooling water systems. The primary reason for supplying low conductivity water to the DIII-D vacuum vessel coils, power supplies and auxiliary heating components is to assure, along with the use of a non-conducting break in the supply piping, sufficient electrical resistance and thus an acceptable current-leakage path to ground at operating voltage potentials. As important, good quality cooling water significantly reduces the likelihood of scaling and fouling of flow passages and heat transfer surfaces. Dissolved oxygen gas removal is also required in one major DIII-D cooling water system to minimize corrosion in the ion sources of the neutral beam injectors. Currently, the combined pumping capacity of the high quality cooling water systems at DIII-D is ∼5,000 gpm. Another area that receives close attention at DIII-D is the chemical treatment of the water used in the cooling towers. This paper discusses the DIII-D water quality requirements, the means used to obtain the necessary quality and the instrumentation used for control and monitoring. Costs to mechanically and chemically condition and maintain water quality are discussed as well as the various aspects of complying with government standards and regulations

Thermohydraulic relationships for advanced water cooled reactors

International Nuclear Information System (INIS)

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Thermohydraulic relationships for advanced water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-04-01

This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

Neutronic design for a 100MW{sub th} Small modular natural circulation lead or lead-alloy cooled fast reactors core

Energy Technology Data Exchange (ETDEWEB)

Chen, C.; Chen, H.; Zhang, H.; Chen, Z.; Zeng, Q., E-mail: shchshch@ustc.edu.cn, E-mail: hlchen1@ustc.edu.cn, E-mail: kulah@mail.ustc.edu.cn, E-mail: zchen214@mail.ustc.edu.cn, E-mail: zengqin@ustc.edu.cn [Univ. of Science and Technology of China, School of Nuclear Science and Technology, Hefei, Anhui (China)

2015-07-01

Lead or lead-alloy cooled fast reactor with good fuel proliferation and nuclear waste transmutation capability, as well as high security and economy, is a great potential for the development of fourth-generation nuclear energy systems. Small natural circulation reactor is an important technical route lead cooled fast reactors industrial applications, which has been chosen as one of the three reference technical for solution lead or lead-alloy cooled fast reactors by GIF lead-cooled fast reactor steering committee. The School of Nuclear Science and Technology of USTC proposed a small 100MW{sub th} natural circulation lead cooled fast reactor concept called SNCLFR-100 based realistic technology. This article describes the SNCLFR-100 reactor of the overall technical program, core physics calculation and analysis. The results show that: SNCLFR-100 with good neutronic and safety performance and relevant design parameters meet the security requirements with feasibility. (author)

Operational cost minimization in cooling water systems

Directory of Open Access Journals (Sweden)

Castro M.M.

2000-01-01

Full Text Available In this work, an optimization model that considers thermal and hydraulic interactions is developed for a cooling water system. It is a closed loop consisting of a cooling tower unit, circulation pump, blower and heat exchanger-pipe network. Aside from process disturbances, climatic fluctuations are considered. Model constraints include relations concerning tower performance, air flowrate requirement, make-up flowrate, circulating pump performance, heat load in each cooler, pressure drop constraints and climatic conditions. The objective function is operating cost minimization. Optimization variables are air flowrate, forced water withdrawal upstream the tower, and valve adjustment in each branch. It is found that the most significant operating cost is related to electricity. However, for cooled water temperatures lower than a specific target, there must be a forced withdrawal of circulating water and further makeup to enhance the cooling tower capacity. Additionally, the system is optimized along the months. The results corroborate the fact that the most important variable on cooling tower performance is not the air temperature itself, but its humidity.

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

2015-02-15

Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

Directory of Open Access Journals (Sweden)

Yong Kyo Lee

2015-02-01

Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

Modern precise high-power water-cooling systems for press quenching

Directory of Open Access Journals (Sweden)

A. Patejuk

2009-04-01

Full Text Available Demand for extrusions in transport applications is increasing rapidly. The extrusions must be strong, light, crashworthy and may have to undergo hydroforming. This implies low wall thicknesses (1-2½ mm in strong alloys that need very fast quenching to obtain the required T4 temper. Crashworth iness – the ability to absorb a lot of energy in crushing deformation – demands very uniform properties throughout the section, and so does hydroforming. Various systems of water or air/water jets, with and without scanning, with and withoutarrangements for precisely aiming the jets, have proved effective for less difficult alloys in wall thicknesses down to 3 mm. These areunsuitable for the new types of transport extrusions, either inducing physical distortion or non-uniform mechanical properties. A novelcooling system that satisfies the new requirements uses laminar water jets of 50-250 μm diameter in a densely packed array of up to10/cm2. These are arranged in modules whose position and direction of aim can be adjusted relative to the part of the extrusion they cool,assuring linear cooling of all parts of the section at up to 500 K/s. The array of modules is very compact and not expensive. A sophisticated system of water microfiltration ensures that the fine nozzles do not become blocked.

Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

International Nuclear Information System (INIS)

Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

1979-01-01

Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

2014-10-15

Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

WGOTHIC analysis of AP1000 passive containment cooling water

International Nuclear Information System (INIS)

Ye Cheng; Wang Yong; Zheng Mingguang; Wang Guodong; Zhang Di; Ni Chenxiao; Wang Minglu

2013-01-01

The WGOTHIC code was used to analyze the influence of the containment cooling water inventory to containment safety for different cases. The results show that if passive containment cooling system fails, the pressure in containment is beyond design limit after 1000 s; if cooling water can't be supplied after 72 h, the pressure in containment is beyond design limit after 0.9 d; if cooling water can't be supplied after 19.6 d, the pressure in containment is beyond design limit but less than the breakdown pressure; if cooling water is supplied for 30 d, the air cooling can remove the decay heat without any aid. It is a reference for making emergency plan and improving containment design. (authors)

A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

Gas cooled fast reactor research in Europe

International Nuclear Information System (INIS)

Stainsby, Richard; Peers, Karen; Mitchell, Colin; Poette, Christian; Mikityuk, Konstantin; Somers, Joe

2011-01-01

Research on the gas-cooled fast reactor system is directed towards fulfilling the ambitious long term goals of Generation IV (Gen IV), i.e., to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. In common with other fast reactors, gas-cooled fast reactors (GFRs) have exceptional potential as sustainable energy sources, for both the utilisation of fissile material and minimisation of nuclear waste through transmutation of minor actinides. The primary goal of GFR research is to develop the system primarily to be a reliable and economic electricity generator, with good safety and sustainability characteristics. However, for the longer term, GFR retains the potential for hydrogen production and other process heat applications facilitated through a high core outlet temperature which, in this case, is not limited by the characteristics of the coolant. In this respect, GFR can inherit the non-electricity applications of the thermal HTRs in a sustainable manner in a future in which natural uranium becomes scarce. GFR research within Europe is performed directly by those states who have signed the 'System Arrangement' document within the Generation IV International Forum (the GIF), specifically France and Switzerland and Euratom. Importantly, Euratom provides a route by which researchers in other European states, and other non-European affiliates, can contribute to the work of the GIF, even when these states are not signatories to the GFR System Arrangement in their own right. This paper is written from the perspective of Euratom's involvement in research on the GFR system, starting with the 5th Framework Programme (FP5) GCFR project in 2000, through the FP6 project between 2005 and 2009 and looking ahead to the proposed activities within the current 7th Framework Programme (FP7). The evolution of the GFR concept from the 1960s onwards is discussed briefly, followed by the current perceived role, objectives and progress with

Liquid metal cooled fast breeder nuclear reactor

International Nuclear Information System (INIS)

Scott, D.

1979-01-01

A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

Deuterium- and 18O-content in the cooling water of power station cooling towers

International Nuclear Information System (INIS)

Heimbach, H.; Dongmann, G.

1976-09-01

The 0-18/0-16 and D/H isotope ratios of water from two different cooling towers were determined by mass spectrometry. The observed isotope fractionation corresponds to that known from natural evaporation or transpiration processes: cooling tower I: delta(D) = 46.8 per thousand, delta( 18 O) = 7.6 per thousand cooling tower II: delta(D) = 33.9 per thousand delta( 18 O) = 5.7 per thousand Evaluation of simple compartment models of a cooling tower and a distillation device suggests that there exists some isotope discrimination within the open trickling unit of a cooling tower analogous to that in a rectification column. In a real cooling tower, however, this effect is compensated largely by the recycling of the cooling water, resulting only in a small enrichment of the heavy isotopes. This can be understood as the result of three partial effects: 1) a fractionation in the vapor pressure equilibrium, 2) a kinetic effect due to diffusion of the water vapor into a turbulent atmosphere, and 3) an exchange effect which is proportional to relative humidity. This low enrichment of the heavy isotope excludes the technical use of cooling towers as isotope separation devices. (orig.) [de

A resting bottom sodium cooled fast reactor

International Nuclear Information System (INIS)

Costes, D.

2012-01-01

This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

Gas cooled fast breeder reactors using mixed carbide fuel

International Nuclear Information System (INIS)

Kypreos, S.

1976-09-01

The fast reactors being developed at the present time use mixed oxide fuel, stainless-steel cladding and liquid sodium as coolant (LMFBR). Theoretical and experimental designing work has also been done in the field of gas-cooled fast breeder reactors. The more advanced carbide fuel offers greater potential for developing fuel systems with doubling times in the range of ten years. The thermohydraulic and physics performance of a GCFR utilising this fuel is assessed. One question to be answered is whether helium is an efficient coolant to be coupled with the carbide fuel while preserving its superior neutronic performance. Also, an assessment of the fuel cycle cost in comparison to oxide fuel is presented. (Auth.)

Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

International Nuclear Information System (INIS)

Yoshiaki Oka; Sei-ichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji

2002-01-01

The paper describes elements of design consideration of supercritical-pressure, light water cooled reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water cooled reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential of producing energy products such as hydrogen and high quality hydro carbons. (authors)

Linear Dynamics Model for Steam Cooled Fast Power Reactors

Energy Technology Data Exchange (ETDEWEB)

Vollmer, H

1968-04-15

A linear analytical dynamic model is developed for steam cooled fast power reactors. All main components of such a plant are investigated on a general though relatively simple basis. The model is distributed in those parts concerning the core but lumped as to the external plant components. Coolant is considered as compressible and treated by the actual steam law. Combined use of analogue and digital computer seems most attractive.

Linear programming optimization of nuclear energy strategy with sodium-cooled fast reactors

International Nuclear Information System (INIS)

Lee, Je Whan; Jeong, Yong Hoon; Chang, Yoon Il; Chang, Soon Heung

2011-01-01

Nuclear power has become an essential part of electricity generation to meet the continuous growth of electricity demand. A Sodium-cooled Fast Reactor (SFR) was developed to extend uranium resource utilization under a growing nuclear energy scenario while concomitantly providing a nuclear waste management solution. Key questions in this scenario are when to introduce SFRs and how many reactors should be introduced. In this study, a methodology using Linear Programming is employed in order to quantify an optimized growth pattern of a nuclear energy system comprising light water reactors and SFRs. The optimization involves tradeoffs between SFR capital cost premiums and the total system U3O8 price premiums. Optimum nuclear growth patterns for several scenarios are presented, as well as sensitivity analyses of important input parameters

Cooling of gas turbines IX : cooling effects from use of ceramic coatings on water-cooled turbine blades

Science.gov (United States)

Brown, W Byron; Livingood, John N B

1948-01-01

The hottest part of a turbine blade is likely to be the trailing portion. When the blades are cooled and when water is used as the coolant, the cooling passages are placed as close as possible to the trailing edge in order to cool this portion. In some cases, however, the trailing portion of the blade is so narrow, for aerodynamic reasons, that water passages cannot be located very near the trailing edge. Because ceramic coatings offer the possibility of protection for the trailing part of such narrow blades, a theoretical study has been made of the cooling effect of a ceramic coating on: (1) the blade-metal temperature when the gas temperature is unchanged, and (2) the gas temperature when the metal temperature is unchanged. Comparison is also made between the changes in the blade or gas temperatures produced by ceramic coatings and the changes produced by moving the cooling passages nearer the trailing edge. This comparison was made to provide a standard for evaluating the gains obtainable with ceramic coatings as compared to those obtainable by constructing the turbine blade in such a manner that water passages could be located very near the trailing edge.

Present status and future perspective of R and D on lead heavy metal-cooled fast reactors

International Nuclear Information System (INIS)

Takahashi, Minoru

2007-01-01

Since a lead heavy metal (lead-bismuth eutectic) is chemically inert and has higher boiling point compared to a sodium, a lead heavy metal-cooled fast reactor can be inherently safe and has good nuclear characteristics and is so suitable to a medium-small size of the reactor. R and D on corrosion of a lead heavy metal has been carried out in the world and this issue might be solved to choose specific corrosion resistant alloys for structural materials and fuel cans of a lead heavy metal-cooled reactor. This article reviews present status and future perspective on lead heavy metal-cooled fast reactors. (T. Tanaka)

Hydrogen migration in fast cooled Pd-H alloys around 50 K

International Nuclear Information System (INIS)

Yamakawa, Kohji

2005-01-01

The migration of hydrogen in fast cooled Pd-H alloys is investigated by electrical resistivity measurement around 50 K. The disordered atoms of hydrogen are introduced by fast cooling from around 80 K to around 20 K. The disordered atoms are ordered by migration of the atoms during heating-up of the specimens with the constant rates. The resistivity is measured at the temperatures as done by Vajda et al. to compare our results with their results. The activation energy of migration of hydrogen is obtained from the kinetic analysis of the resistivity change due to the ordering using a so-called cross-cut method. The obtained value is compared with our previous results, which have been obtained with hydrogen disordered by quenching into liquid helium, isothermal annealing of the specimen and resistivity measurement in liquid helium. The present value is in fairly good agreement with the value obtained in the quenched case

Conceptual core designs for a 1200 MWe sodium cooled fast reactor

International Nuclear Information System (INIS)

Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

2008-01-01

The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

On synthesis and optimization of cooling water systems with multiple cooling towers

CSIR Research Space (South Africa)

Gololo, KV

2011-01-01

Full Text Available -1 On Synthesis and Optimization of Cooling Water Systems with Multiple Cooling Towers Khunedi Vincent Gololo?? and Thokozani Majozi*? ? Department of Chemical Engineering, University of Pretoria, Lynnwood Road, Pretoria, 0002, South Africa ? Modelling...

Cooling water for SSC experiments: Supplemental Conceptual Design Report (SCDR)

International Nuclear Information System (INIS)

Doyle, R.E.

1989-01-01

This paper discusses the following topics on cooling water design on the superconducting super collider; low conductivity water; industrial cooling water; chilled water systems; and radioactive water systems

Cooling water conditioning and quality control for tokamaks

International Nuclear Information System (INIS)

Gootgeld, A.M.

1995-10-01

Designers and operators of Tokamaks and all associated water cooled, peripheral equipment, are faced with the task of providing and maintaining closed-loop, low conductivity, low impurity, cooling water systems. Most of these systems must provide large volumes of high quality cooling water at reasonable cost and comply with local and state government orders and EPA mandated national pretreatment standards and regulations. This paper discusses the DIII-D water quality requirements, the means used to obtain the necessary quality and the instrumentation used for control and monitoring. Costs to mechanically and chemically condition and maintain water quality are discussed as well as the various aspects of complying with government standards and regulations

Closed cooling water chemistry guidelines revision

International Nuclear Information System (INIS)

McElrath, Joel; Breckenridge, Richard

2014-01-01

This second revision of the Closed Cooling Water Chemistry Guideline addresses the use of chemicals and monitoring methods to mitigate corrosion, fouling, and microbiological growth in the closed cooling-water (CCW) systems of nuclear and fossil-fueled power plants. This revision has been endorsed by the utility chemistry community and represents another step in developing a more proactive chemistry program to limit or control closed cooling system degradation with increased consideration of corporate resources and plant-specific design and operating concerns. These guidelines were developed using laboratory data, operating experience, and input from organizations and utilities within and outside of the United States of America. It is the intent of the Revision Committee that these guidelines are applicable to all nuclear and fossil-fueled generating stations around the world. A committee of industry experts—including utility specialists, Institute of Nuclear Power Operations representatives, water-treatment service-company representatives, consultants, a primary contractor, and EPRI staff—collaborated in reviewing available data on closed cooling-water system corrosion and microbiological issues. Recognizing that each plant owner has a unique set of design, operating, and corporate concerns, the Guidelines Committee developed a methodology for plant-specific optimization. The guideline provides the technical basis for a reasonable but conservative set of chemical treatment and monitoring programs. The use of operating ranges for the various treatment chemicals discussed in this guideline will allow a power plant to limit corrosion, fouling, and microbiological growth in CCW systems to acceptable levels. The guideline now includes closed cooling chemistry regimes proven successful in use in the international community. The guideline provides chemistry constraints for the use of phosphates control, as well as pure water with pH control. (author)

CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

Directory of Open Access Journals (Sweden)

Pengcheng Zhao

2016-01-01

Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

Membrane distillation of industrial cooling tower blowdown water

Directory of Open Access Journals (Sweden)

N.E. Koeman-Stein

2016-06-01

Full Text Available The potential of membrane distillation for desalination of cooling tower blowdown water (CTBD is investigated. Technical feasibility is tested on laboratory and pilot scale using real cooling tower blowdown water from Dow Benelux in Terneuzen (Netherlands. Two types of membranes, polytetrafluorethylene and polyethylene showed good performance regarding distillate quality and fouling behavior. Concentrating CTBD by a factor 4.5 while maintaining a flux of around 2 l/m2*h was possible with a water recovery of 78% available for reuse. Higher concentration factors lead to severe decrease in flux which was caused by scaling. Membrane distillation could use the thermal energy that would otherwise be discharged of in a cooling tower and function as a heat exchanger. This reduces the need for cooling capacity and could lead to a total reduction of 37% water intake for make-up water, as well as reduced energy and chemicals demands and greenhouse gas emissions.

The Water Quality Control of the Secondary Cooling Water under a Normal Operation of 30 MWth in HANARO

International Nuclear Information System (INIS)

Park, Young Chul; Lee, Young Sub; Lim, Rag Yong

2008-01-01

HANARO, a multi-purpose research reactor, a 30 MWth open-tank-in-pool type, has been under a full power operation since 2005. The heat generated by the core of HANARO is transferred to the primary cooling water. And the cooling water transfers the heat to the secondary cooling water through the primary cooling heat exchanger. The heat absorbed by the secondary cooling water is removed through a cooling tower. The quality of the secondary cooling water is deteriorated by a temperature variation of the cooling water and a foreign material flowing over the cooling water through the cooling tower fan for a cooling. From these, a corrosion reduces the life time of a system, a scale degrades the heat transfer effect and a sludge and slime induces a local corrosion. For reducing these impacts, the quality of the secondary cooling water is treated by a high ca-hardness water quality program by maintaining a super saturated condition of ions, 12 of a ca-hardness concentration. After an overhaul maintenance of a secondary cooling tower composed of a secondary cooling system in 2007, a secondary cooling water stored in the cooling tower basin was replaced with a fresh city water. In this year, a water quality deterioration test has been performed under a full power operation and a mode of a twenty three day operation and twelve day maintenance for setting a beginning control limit of the secondary cooling water. This paper describes the water quality deterioration test for the secondary cooling system under a full power operation of 30 MWth including a test method, a test requirement and a test result

Heavy water moderated gas-cooled reactors

International Nuclear Information System (INIS)

Bailly du Bois, B.; Bernard, J.L.; Naudet, R.; Roche, R.

1964-01-01

France has based its main effort for the production of nuclear energy on natural Uranium Graphite-moderated gas-cooled reactors, and has a long term programme for fast reactors, but this country is also engaged in the development of heavy water moderated gas-cooled reactors which appear to present the best middle term prospects. The economy of these reactors, as in the case of Graphite, arises from the use of natural or very slightly enriched Uranium; heavy water can take the best advantages of this fuel cycle and moreover offers considerable development potential because of better reactor performances. A prototype plant EL 4 (70 MW) is under construction and is described in detail in another paper. The present one deals with the programme devoted to the development of this reactor type in France. Reasons for selecting this reactor type are given in the first part: advantages and difficulties are underlined. After reviewing the main technological problems and the Research and Development carried out, results already obtained and points still to be confirmed are reported. The construction of EL 4 is an important step of this programme: it will be a significant demonstration of reactor performances and will afford many experimentation opportunities. Now the design of large power reactors is to be considered. Extension and improvements of the mechanical structures used for EL 4 are under study, as well as alternative concepts. The paper gives some data for a large reactor in the present state of technology, as a result from optimization studies. Technical improvements, especially in the field of materials could lead to even more interesting performances. Some prospects are mentioned for the long run. Investment costs and fuel cycles are discussed in the last part. (authors) [fr

Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

Water supply method to the fuel cell cooling water system; Nenryo denchi reikyakusuikei eno kyusui hoho

Energy Technology Data Exchange (ETDEWEB)

Urata, T. [Tokyo (Japan); Nishida, S. [Tokyo (Japan)

1996-12-17

The conventional fuel cell has long cooling water piping ranging from the fuel cell exit to the steam separator; in addition, the supply water is cooler than the cooling water. When the amount of supply water increases, the temperature of the cooling water is lowered, and the pressure fluctuation in the steam separator becomes larger. This invention relates to the water supply method of opening the supply water valve and supplying water from the supply water system to the cooling water system in accordance with the signal of the level sensor of the steam separator, wherein opening and closing of the supply valve are repeated during water supply. According to the method the pressure drop in every water supply becomes negligibly small; therefore, the pressure fluctuation of the cooling water system can be made small. The interval of the supply water valve from opening to closing is preferably from 3 seconds to 2 minutes. The method is effective when equipment for recovering heat from the cooling water is installed in the downstream pipeline of the fuel cell. 2 figs.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

State waste discharge permit application 400 Area secondary cooling water. Revision 2

International Nuclear Information System (INIS)

1996-01-01

This document constitutes the Washington Administrative Code 173-216 State Waste Discharge Permit Application that serves as interim compliance as required by Consent Order DE 91NM-177, for the 400 Area Secondary Cooling Water stream. As part of the Hanford Federal Facility Agreement and Consent Order negotiations, the US Department of Energy, Richland Operations Office, the US Environmental Protection Agency, and the Washington State Department of Ecology agreed that liquid effluent discharges to the ground on the Hanford Site that affect groundwater or have the potential to affect groundwater would be subject to permitting under the structure of Chapter 173-216 of the Washington Administrative Code, the State Waste Discharge Permitting Program. As a result of this decision, the Washington State Department of Ecology and the US Department of Energy, Richland Operations Office entered into Consent Order DE 91NM-177. The Consent Order DE 91NM-177 requires a series of permitting activities for liquid effluent discharges. Based upon compositional and flow rate characteristics, liquid effluent streams on the Hanford Site have been categorized into Phase 1, Phase 2, and Miscellaneous streams. This document only addresses the 400 Area Secondary Cooling Water stream, which has been identified as a Phase 2 stream. The 400 Area Secondary Cooling Water stream includes contribution streams from the Fuels and Materials Examination Facility, the Maintenance and Storage Facility, the 481-A pump house, and the Fast Flux Test Facility

Future work in the DeBeNeLux research centres on the sodium cooled fast reactor

International Nuclear Information System (INIS)

Goedkoop, J.A.

1976-01-01

The general objectives as they now apply over the world in the further development of the sodium cooled fast reactor are to realize a reactor and the associated fuel cycle, that will ensure a good fuel utilization; secondly, as long as we live in a more or less free market economy, such a system will only be acceptable if it is competitive, which means that the difference in investment cost between the fast reactor and the presently used light water reactors has to be brought down; thirdly, to justify the investment the system should work reliably; finally the developments in reactor design should not be at the expense of reactor safety. The pursuit of these objectives during the coming years will require the DeBeNeLuX laboratories to do work in a number of fields. (Auth.)

The atmospheric cooling of nuclear power stations

International Nuclear Information System (INIS)

Leuenberger, J.M.; Mayor, J.C.; Gassmann, F.; Lieber, K.

1978-08-01

Four different types of nuclear reactor are considered: light water reactors, high temperature reactors with steam circulation and with direct gas turbine circulation, and fast breeder reactors. Wet and dry cooling towers are described and experimental studies carried out using cooling tower models are presented. (G.T.H.)

A model for radionuclide transport in the Cooling Water System

International Nuclear Information System (INIS)

Kahook, S.D.

1992-08-01

A radionuclide transport model developed to assess radiological levels in the K-reactor Cooling Water System (CWS) in the event of an inadvertent process water (PW) leakage to the cooling water (CW) in the heat exchangers (HX) is described. During and following a process water leak, the radionuclide transport model determines the time-dependent release rates of radionuclide from the cooling water system to the environment via evaporation to the atmosphere and blow-down to the Savannah River. The developed model allows for delay times associated with the transport of the cooling water radioactivity through cooling water system components. Additionally, this model simulates the time-dependent behavior of radionuclides levels in various CWS components. The developed model is incorporated into the K-reactor Cooling Tower Activity (KCTA) code. KCTA allows the accident (heat exchanger leak rate) and the cooling tower blow-down and evaporation rates to be described as time-dependent functions. Thus, the postulated leak and the consequence of the assumed leak can be modelled realistically. This model is the first of three models to be ultimately assembled to form a comprehensive Liquid Pathway Activity System (LPAS). LPAS will offer integrated formation, transport, deposition, and release estimates for radionuclides formed in a SRS facility. Process water and river water modules are forthcoming as input and downstream components, respectively, for KCTA

Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply

International Nuclear Information System (INIS)

Sienicki, J.J.; Wade, D.C.; Moisseytsev, A.

2008-01-01

Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus

Corrosion induced clogging and plugging in water-cooled generator cooling circuit

International Nuclear Information System (INIS)

Park, B.G.; Hwang, I.S.; Rhee, I.H.; Kim, K.T.; Chung, H.S.

2002-01-01

Water-cooled electrical generators have been experienced corrosion-related problems that are restriction of flow through water strainers caused by collection of excessive amounts of copper corrosion products (''clogging''), and restriction of flow through the copper strands in the stator bars caused by growth or deposition of corrosion products on the walls of the hollow strands (''plugging''). These phenomena result in unscheduled shutdowns that would be a major concern because of the associated loss in generating capacity. Water-cooled generators are operated in one of two modes. They are cooled either with aerated water (dissolved oxygen >2 ppm) or with deaerated water (dissolved oxygen <50 ppb). Both modes maintain corrosion rates at satisfactorily low levels as long as the correct oxygen concentrations are maintained. However, it is generally believed that very much higher copper corrosion rates result at the intermediate oxygen concentrations of 100-1000 ppb. Clogging and plugging are thought to be associated with these intermediate concentrations, and many operators have suggested that the period of change from high-to-low or from low-to-high oxygen concentration is particularly damaging. In order to understand the detailed mechanism(s) of the copper oxide formation, release and deposition and to identify susceptible conditions in the domain of operating variables, a large-scale experiments are conducted using six hollow strands of full length connected with physico-chemically scaled generator cooling water circuit. To ensure a close simulation of thermal-hydraulic conditions in a generator stator, strands of the loop will be ohmically heated using AC power supply. Experiments is conducted to cover oxygen excursions in both high dissolved oxygen and low dissolved oxygen conditions that correspond to two representative operating condition at fields. A thermal upset condition is also simulated to examine the impact of thermal stress. During experiments

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Duncombe, E.; Thatcher, G.

1979-01-01

The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

Design and selection of materials for sodium cooled fast reactors

International Nuclear Information System (INIS)

Chetal, S.C.

2011-01-01

Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

Study on plant concept for gas cooled fast reactor

International Nuclear Information System (INIS)

Moribe, Takeshi; Kubo, Shigenobu; Saigusa, Toshiie; Konomura, Mamoru

2003-05-01

In 'Feasibility Study on Commercialized Fast Reactor Cycle System', technological options including various coolant (sodium, heavy metal, gas, water, etc.), fuel type (MOX, metal, nitride) and output power are considered and classified, and commercialized FBR that have economical cost equal to LWR are pursued. In conceptual study on gas cooled FBR in FY 2002, to identify the prospect of the technical materialization of the helium cooled FBR using coated particle fuel which is an attractive concept extracted in the year of FY2001, the preliminary conceptual design of the core and entire plant was performed. This report summarizes the results of the plant design study in FY2002. The results of study is as follows. 1) For the passive core shutdown equipment, the curie point magnet type self-actuated device was selected and the device concept was set up. 2) For the reactor block, the concept of the core supporting structure, insulators and liners was set up. For the material of the heat resistant structure, SiC was selected as a candidate. 3) For the seismic design of the plant, it was identified that a design concept with three-dimensional base isolation could be feasible taking the severe seismic condition into account. 4) For the core catcher, an estimation of possible event sequences under severe core damage condition was made. A core catcher concept which may suit the estimation was proposed. 5) The construction cost was roughly estimated based on the amount of materials and its dependency on the plant output power was evaluated. The value for a small sized plant exceeds the target construction cost about 20%. (author)

Influence of detergents on water drift in cooling towers

Science.gov (United States)

Vitkovicova, Rut

An influence of detergents on the water drift from the cooling tower was experimentally investigated. For this experimental measurements was used a model cooling tower, especially an experimental aerodynamic line, which is specially designed for the measurement and monitoring of processes taking place around the eliminators of the liquid phase. The effect of different concentrations of detergent in the cooling water on the drift of water droplets from a commonly used type eliminator was observed with visualization methods.

Engineering the mobility increment in pentacene-based field-effect transistors by fast cooling of polymeric modification layer

Science.gov (United States)

Ling, Haifeng; Zhang, Chenxi; Chen, Yan; Shao, Yaqing; Li, Wen; Li, Huanqun; Chen, Xudong; Yi, Mingdong; Xie, Linghai; Huang, Wei

2017-06-01

In this work, we investigate the effect of the cooling rate of polymeric modification layers (PMLs) on the mobility improvement of pentacene-based organic field-effect transistors (OFETs). In contrast to slow cooling (SC), the OFETs fabricated through fast cooling (FC) with PMLs containing side chain-phenyl rings, such as polystyrene (PS) and poly (4-vinylphenol) (PVP), show an obvious mobility incensement compared with that of π-group free polymethylmethacrylate (PMMA). Atomic force microscopy (AFM) images and x-ray diffraction (XRD) characterizations have showed that fast-cooled PMLs could effectively enhance the crystallinity of pentacene, which might be related to the optimized homogeneity of surface energy on the surface of polymeric dielectrics. Our work has demonstrated that FC treatment could be a potential strategy for performance modulation of OFETs.

Wetland Water Cooling Partnership: The Use of Constructed Wetlands to Enhance Thermoelectric Power Plant Cooling and Mitigate the Demand of Surface Water Use

Energy Technology Data Exchange (ETDEWEB)

Apfelbaum, Steven L. [Applied Ecological Services Inc., Brodhead, WI (United States); Duvall, Kenneth W. [Sterling Energy Services, LLC, Atlanta, GA (United States); Nelson, Theresa M. [Applied Ecological Services Inc., Brodhead, WI (United States); Mensing, Douglas M. [Applied Ecological Services Inc., Brodhead, WI (United States); Bengtson, Harlan H. [Sterling Energy Services, LLC, Atlanta, GA (United States); Eppich, John [Waterflow Consultants, Champaign, IL (United States); Penhallegon, Clayton [Sterling Energy Services, LLC, Atlanta, GA (United States); Thompson, Ry L. [Applied Ecological Services Inc., Brodhead, WI (United States)

2013-12-01

Through the Phase I study segment of contract #DE-NT0006644 with the U.S. Department of Energy’s National Energy Technology Laboratory, Applied Ecological Services, Inc. and Sterling Energy Services, LLC (the AES/SES Team) explored the use of constructed wetlands to help address stresses on surface water and groundwater resources from thermoelectric power plant cooling and makeup water requirements. The project objectives were crafted to explore and develop implementable water conservation and cooling strategies using constructed wetlands (not existing, naturally occurring wetlands), with the goal of determining if this strategy has the potential to reduce surface water and groundwater withdrawals of thermoelectric power plants throughout the country. Our team’s exploratory work has documented what appears to be a significant and practical potential for augmenting power plant cooling water resources for makeup supply at many, but not all, thermoelectric power plant sites. The intent is to help alleviate stress on existing surface water and groundwater resources through harvesting, storing, polishing and beneficially re-using critical water resources. Through literature review, development of conceptual created wetland plans, and STELLA-based modeling, the AES/SES team has developed heat and water balances for conventional thermoelectric power plants to evaluate wetland size requirements, water use, and comparative cooling technology costs. The ecological literature on organism tolerances to heated waters was used to understand the range of ecological outcomes achievable in created wetlands. This study suggests that wetlands and water harvesting can provide a practical and cost-effective strategy to augment cooling waters for thermoelectric power plants in many geographic settings of the United States, particularly east of the 100th meridian, and in coastal and riverine locations. The study concluded that constructed wetlands can have significant positive

Influence of detergents on water drift in cooling towers

Directory of Open Access Journals (Sweden)

Vitkovicova Rut

2017-01-01

Full Text Available An influence of detergents on the water drift from the cooling tower was experimentally investigated. For this experimental measurements was used a model cooling tower, especially an experimental aerodynamic line, which is specially designed for the measurement and monitoring of processes taking place around the eliminators of the liquid phase. The effect of different concentrations of detergent in the cooling water on the drift of water droplets from a commonly used type eliminator was observed with visualization methods.

Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

International Nuclear Information System (INIS)

Tomabechi, K.

1975-01-01

The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

International Nuclear Information System (INIS)

Park, Jee Won; Jeong, C. J.; Yang, M. S.

2003-03-01

Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

Energy Technology Data Exchange (ETDEWEB)

Park, Jee Won; Jeong, C. J.; Yang, M. S

2003-03-01

Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

Effect of closed loop cooling water transit time on containment cooling

International Nuclear Information System (INIS)

Smith, R.P.; Vossahlik, J.E.; Goodwin, E.F.

1996-01-01

Long term containment cooling analyses in nuclear plant systems are usually conducted assuming a quasi steady-state process, that is, a steady state evaluation of the cooling system is completed for each calculational step. In reality, fluid transport in the system, and heat addition to system components may affect the heat removal rate of the system. Transient effects occurring during system startup may affect the maximum temperatures experienced in the system. It is important to ensure that such transient effects do not affect operation of the system (e.g., cause a high temperature trip). To evaluate the effect of fluid transit delays, a closed loop cooling water system model has been developed that incorporates the fluid transport times when determining the closed loop cooling system performance. This paper describes the closed loop cooling system model as implemented in the CONTEMPT-LT/028 code. The evaluation of the transient temperature response of the closed loop cooling system using the model is described. The paper also describes the effect of fluid transit time on the overall containment cooling performance

Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-08-15

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Sang June Ahn

2016-08-01

Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

Energy Technology Data Exchange (ETDEWEB)

Gary Vine

2010-12-01

This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes “Best Technology Available” for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant’s steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R&D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

COOLING WATER ISSUES AND OPPORTUNITIES AT U.S. NUCLEAR POWER PLANTS

International Nuclear Information System (INIS)

Vine, Gary

2010-01-01

This report has been prepared for the Department of Energy, Office of Nuclear Energy (DOE-NE), for the purpose of providing a status report on the challenges and opportunities facing the U.S. commercial nuclear energy industry in the area of plant cooling water supply. The report was prompted in part by recent Second Circuit and Supreme Court decisions regarding cooling water system designs at existing thermo-electric power generating facilities in the U.S. (primarily fossil and nuclear plants). At issue in the courts have been Environmental Protection Agency regulations that define what constitutes 'Best Technology Available' for intake structures that withdraw cooling water that is used to transfer and reject heat from the plant's steam turbine via cooling water systems, while minimizing environmental impacts on aquatic life in nearby water bodies used to supply that cooling water. The report was also prompted by a growing recognition that cooling water availability and societal use conflicts are emerging as strategic energy and environmental issues, and that research and development (R and D) solutions to emerging water shortage issues are needed. In particular, cooling water availability is an important consideration in siting decisions for new nuclear power plants, and is an under-acknowledged issue in evaluating the pros and cons of retrofitting cooling towers at existing nuclear plants. Because of the significant ongoing research on water issues already being performed by industry, the national laboratories and other entities, this report relies heavily on ongoing work. In particular, this report has relied on collaboration with the Electric Power Research Institute (EPRI), including its recent work in the area of EPA regulations governing intake structures in thermoelectric cooling water systems.

Energy and water management in evaporative cooling systems in Saudi Arabia

Energy Technology Data Exchange (ETDEWEB)

Kassem, Abdel-wahab S. (Agricultural and Veterinary Training and Research Station, King Faisal University, Al-Hassa (Saudi Arabia))

1994-11-01

A mathematical model was developed to estimate water evaporation rate, airflow rate and cooling effect in an evaporative cooling system for farm structures. The model was only applied to evaporative cooling systems for greenhouses. The effect of ambient air temperature, solar radiation and system efficiency on water evaporation rate, airflow rate and the resulting cooling effect were studied. Generally, water flow rate and air flow rate are adjusted based on daily maximum temperature. However, a substantial saving in energy and water consumption in the cooling system would be achieved by regulating water flow rate and air flow rate to follow the diurnal variation on temperature. Improving the cooling efficiency and covering the roof of the greenhouse with an external shading would save an appreciable amount of energy and water consumption. The model could also be applied to other farm structures such as animal shelters

Environmental compatible cooling water treatment chemicals; Umweltvertraegliche Chemikalien in der Kuehlwasserkonditionierung

Energy Technology Data Exchange (ETDEWEB)

Gartiser, S; Urich, E

2002-02-01

In Germany about 32 billion m{sup 3}/a cooling water are discharged from industrial plants and power industry. These are conditioned partly with biocides, scaling and corrosion inhibitors. Within the research project the significance of cooling water chemicals was evaluated, identifying the chemicals from product information, calculating their loads from consumption data of more than 180 cooling plants and investigating the basic data needed for an environmental hazard assessment. Additionally the effects of cooling water samples and products were determined in biological test systems. Batch tests were performed under defined conditions in order to measure the inactivation of cooling water biocides. (orig.)

Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

Directory of Open Access Journals (Sweden)

Chan Bock Lee

2016-10-01

Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

Research and development of super light water reactors and super fast reactors in Japan

International Nuclear Information System (INIS)

Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

2011-01-01

Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

Refueling system for the gas-cooled fast breeder reactor

International Nuclear Information System (INIS)

Hawke, B.C.

1980-05-01

Criteria specifically related to the handling of Gas-Cooled Fast Breeder Reactor (GCFR) fuel are briefly reviewed, and the most significant requirements with which the refueling system must comply are discussed. Each component of the refueling system is identified, and a functional description of the fuel handling machine is presented. An illustrated operating sequence describing the various functions involved in a typical refueling cycle is presented. The design status of components and subsystems selected for conceptual development is reviewed, and anticipated refueling time frames are given

Water cooling system for sintering furnaces of nuclear fuel pellets; Sistema de enfriamiento con agua para hornos de sinterizado de pastillas de combustible nuclear

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-11-01

This work has as a main objective to develop a continuous cooling water system, which is necessary for the cooling of the sintering furnaces. This system is used to protect them as well as for reducing the water consumption, ejecting the heat generated into this furnaces and scattering it into the atmosphere in a fast and continuous way. The problem was defined and the reference parameters established, making the adequate research. The materials were selected as well as the length of the pipeline which will carry the secondary refrigerant fluid (water). Three possible solutions were tried,and evaluated, and from these, the thermal and economically most efficient option was selected. The layout of the solution was established and the theoretical construction of a cooling system for liquids using dichlorofluoromethane (R-22), as a refrigerant and a air cooled condenser, was accomplished. (Author).

Pilot-scale cooling tower to evaluate corrosion, scaling, and biofouling control strategies for cooling system makeup water.

Science.gov (United States)

Chien, S H; Hsieh, M K; Li, H; Monnell, J; Dzombak, D; Vidic, R

2012-02-01

Pilot-scale cooling towers can be used to evaluate corrosion, scaling, and biofouling control strategies when using particular cooling system makeup water and particular operating conditions. To study the potential for using a number of different impaired waters as makeup water, a pilot-scale system capable of generating 27,000 kJ∕h heat load and maintaining recirculating water flow with a Reynolds number of 1.92 × 10(4) was designed to study these critical processes under conditions that are similar to full-scale systems. The pilot-scale cooling tower was equipped with an automatic makeup water control system, automatic blowdown control system, semi-automatic biocide feeding system, and corrosion, scaling, and biofouling monitoring systems. Observed operational data revealed that the major operating parameters, including temperature change (6.6 °C), cycles of concentration (N = 4.6), water flow velocity (0.66 m∕s), and air mass velocity (3660 kg∕h m(2)), were controlled quite well for an extended period of time (up to 2 months). Overall, the performance of the pilot-scale cooling towers using treated municipal wastewater was shown to be suitable to study critical processes (corrosion, scaling, biofouling) and evaluate cooling water management strategies for makeup waters of complex quality.

Asbestos in cooling-tower waters. Final report

International Nuclear Information System (INIS)

Lewis, B.A.G.

1979-03-01

Water discharges from cooling towers constructed with asbestos fill were found to contain chrysotile--asbestos fibers at concentrations as high as 10 8 fibers/liter. The major source of these fibers, appears to be the components of the towers rather than the air drawn through the towers or the makeup water taken into the towers. Suggested mechanisms for the release of chrysotile fibers from cooling-tower fill include freeze-thaw cycles and dissolution of the cement due to acidic components of the circulating water. Ash- or other material-settling ponds were found to reduce asbestos-fiber concentrations in cooling-tower effluent. The literature reviewed did not support the case for a causal relationship between adverse human health effects and drinking water containing on the order of 10 6 chrysotile--asbestos fibers/liter; for this and other reasons, it is not presently suggested that the use of asbestos fill be discontinued. However, caution and surveillance are dictated by the uncertainties in the epidemiological studies, the absence of evidence for a safe threshold concentration in water, and the conclusive evidence for adverse effects from occupational exposure. It is recommended that monitoring programs be carried out at sites where asbestos fill is used; data from such programs can be used to determine whether any mitigative measures should be taken. On the basis of estimates made in this study, monitoring for asbestos in drift from cooling towers does not appear to be warranted

Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

Energy Technology Data Exchange (ETDEWEB)

Scaller, K; Vrillon, B

1980-02-01

Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component.

Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

International Nuclear Information System (INIS)

Scaller, K.; Vrillon, B.

1980-01-01

Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

Cooling system for the connecting rings of a fast neutron reactor vessel

International Nuclear Information System (INIS)

Martin, J.-P.; Malaval, Claude

1974-01-01

A description is given of a cooling system for the vessel connecting rings of a fast neutron nuclear reactor, particularly of a main vessel containing the core of the reactor and a volume of liquid metal coolant at high temperature and a safety vessel around the main vessel, both vessels being suspended to a rigid upper slab kept at a lower temperature. It is mounted in the annular space between the two vessels and includes a neutral gas circuit set up between the wall of the main vessel to be cooled and that of the safety vessel itself cooled from outer. The neutral gas system comprises a plurality of ventilators fitted in holes made through the thickness of the upper slab and opening on to the space between the two vessels. It also includes two envelopes lining the walls of these vessels, establishing with them small section channels for the circulation of the neutral gas cooled against the safety vessel and heated against the main vessel [fr

Water mist effect on cooling range and efficiency of casting die

Directory of Open Access Journals (Sweden)

R. Władysiak

2008-12-01

Full Text Available This project is showing investigation results of cooling process of casting die in the temperature range 570÷100 °C with 0.40 MPa compressed air and water mist streamed under pressure 0.25÷0.45 MPa in air jet 0.25÷0.50 MPa using open cooling system.The character and the speed of changes of temperature, forming of the temperture’s gradient along parallel layer to cooled surface of die is shawing with thermal and derivative curves. The effect of kind of cooling factor on the temperature and time and distance from cooling nozzle is presented in the paper. A designed device for generating the water mist cooling the die and the view of sprying water stream is shown here. It’s proved that using of the water mist together with the change of heat transfer interface increases intensity of cooling in the zone and makes less the range cooling zone and reduces the porosity of cast microstructure.

Thermal calculations for water cooled research reactors

International Nuclear Information System (INIS)

Fabrega, S.

1979-01-01

The formulae and the more important numerical data necessary for thermic calculations on the core of a research reactor, cooled with low pressure water, are presented. Most of the problems met by the designer and the operator are dealt with (calculations margins, cooling after shut-down). Particular cases are considered (gas release, rough walls, asymmetric cooling slabs etc.), which are not generally envisaged in works on general thermics

R + D work on gas-cooled breeder development

International Nuclear Information System (INIS)

Dalle Donne, M.; Dorner, S.; Jacobs, G.; Meyer, L.; Rehme, K.; Schumacher, G.; Wilhelm, D.

1978-01-01

The development work for the gas-cooled breeder in the Karlsruhe Nuclear Research Center may be assigned to two different groups: a) Investigations on fuel elements. b) Studies concerning the safety of gas-cooled fast breeder reactors. To the first group there belongs the work related to the: - heat transfer between fuel elements and coolant gas, - influence of increased content of water vapor in helium or the fuel rods. The second group concerns: - establishing a computer code for transient calculations in the primary and secondary circuit of a gas-cooled fast breeder reactor, - steam reactivity coefficients, - the core destruction phase of hypothetical accidents, - the core-catcher using borax. (orig./RW) [de

Water-cooled grid ''wires'' for direct converters

International Nuclear Information System (INIS)

Schwer, C.J.

1976-01-01

A study was conducted to determine the feasibility of internal convective cooling of grid ''wires'' for direct converters. Detailed computer calculations reveal that the use of small diameter water cooled tubes as grid ''wires'' is feasible for a considerable range of lengths and thermal fluxes

Accelerator driven light water fast reactor (revisiting to the accelerator LWR fuel regenerator)

International Nuclear Information System (INIS)

Takahashi, H.; Zhang, J.

1999-01-01

A tight-latticed, high-enriched Pu fuel reactor cooled by water or by super-critical steam has a high neutron economy, similar to that of Na-or Pb-cooled fast reactor. Operating in a subcritical condition by providing spallation neutrons, this Pu-fueled reactor can run safely, despite the positive coolant void coefficients. It can be used to transmute the proliferation-prone Pu into proliferation-resistive U-233 fuel using thorium as the fertile material. Rather than employing the large linear accelerator proposed for the LWR fuel regenerator studied in the INFCE program, a small circular accelerator, such as a cyclotron or a Fixed Field Alternating Gradient Synchrotron (FFAG), can run a large power reactor in a slightly subcritical reactor using control rods, on-line fuel reshuffling, and slightly graded proton-beam injection. Some thoughts on improving the reliability of the proton accelerator, on transmutation of the long-lived fission products of Tc-99, and I-129, and the future direction of the development of the fast reactor are discussed. (author)

Instrumentation for NBI SST-1 cooling water system

International Nuclear Information System (INIS)

Qureshi, Karishma; Patel, Paresh; Jana, M.R.

2015-01-01

Neutral Beam Injector (NBI) System is one of the heating systems for Steady state Superconducting Tokamak (SST-1). It is capable of generating a neutral hydrogen beam of power 0.5 MW at 30 kV. NBI system consists of following sub-systems: Ion source, Neutralizer, Deflection Magnet and Magnet Liner (ML), Ion Dump (ID), V-Target (VT), Pre Duct Scraper (PDS), Beam Transmission Duct (BTD) and Shine Through (ST). For better heat removal management purpose all the above sub-systems shall be equipped with Heat Transfer Elements (THE). During beam operation these sub-systems gets heated due to the received heat load which requires to be removed by efficient supplying water. The cooling water system along with the other systems (External Vacuum System, Gas Feed System, Cryogenics System, etc.) will be controlled by NBI Programmable Logic Control (PLC). In this paper instrumentation and its related design for cooling water system is discussed. The work involves flow control valves, transmitters (pressure, temperature and water flow), pH and conductivity meter signals and its interface with the NBI PLC. All the analog input, analog output, digital input and digital output signals from the cooling water system will be isolated and then fed to the NBI PLC. Graphical Users Interface (GUI) needed in the Wonderware SCADA for the cooling water system shall also be discussed. (author)

Strength analysis of fast gas cooled reactor fuel element in conditions of fuel-cladding interraction and non-uniform azimuthal heating

International Nuclear Information System (INIS)

Kulikov, I.S.; Tverkovkin, B.E.

1984-01-01

The technique and the PRORT mathematical program in FORTRAN language for determining mechanical properties of a fuel element with motionless fuel-cladding interaction taking into account circular temperature non-uniformity in gas-cooled fast reactor conditions are proposed. The calculation results of the fuel element of dissociating gas cooled fast reactor are presented for seven cross-sections over the height of the core. The obtained data testify to appreciable swelling of Cr16Ni15Mo3Nb steel fuel cladding in the conditions of dissociating gas cooled fast reactor through the allowance for the effect of stresses on this essential parameter shows, that its value is lower in comparison with swelling, wherein stresses are not taken into account

Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

International Nuclear Information System (INIS)

Kasten, P.R.

1981-09-01

Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies

Analysis of a sustainable gas cooled fast breeder reactor concept

International Nuclear Information System (INIS)

Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

2014-01-01

Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

Design change of tower cooling water system for proton accelerator research center

International Nuclear Information System (INIS)

Jeon, G. P.; Kim, J. Y.; Song, I. T.; Min, Y. S.; Mun, K. J.; Cho, J. S.; Nam, J. M.; Park, S. S.; Han, Y. G.

2012-01-01

The Tower Cooling Water System (TC) is designed to reject the heat load generated by operating the accelerators and the utility facilities through the component cooling water (CCW) heat exchangers. The circulating water discharged from the circulating water pumps passes through the CCW heat exchangers, the Chiller condenser and the air compressor, and the heated circulating water is return to the cooling tower for the heat removal. In this study, The design of Tower Cooling Water System is changed as follows : At First, The quantity of cells is changed into six in order to operate the cooling tower accurately correspond with condition of each equipment of head loads. The fans of cooling tower are controlled by the signal of TEW installed in the latter parts of it. The type of circulation water pump is modified to centrifugal pump and debris filter system is deleted

Design change of tower cooling water system for proton accelerator research center

Energy Technology Data Exchange (ETDEWEB)

Jeon, G. P.; Kim, J. Y.; Song, I. T.; Min, Y. S.; Mun, K. J.; Cho, J. S.; Nam, J. M.; Park, S. S.; Han, Y. G. [KAERI, Daejeon (Korea, Republic of)

2012-10-15

The Tower Cooling Water System (TC) is designed to reject the heat load generated by operating the accelerators and the utility facilities through the component cooling water (CCW) heat exchangers. The circulating water discharged from the circulating water pumps passes through the CCW heat exchangers, the Chiller condenser and the air compressor, and the heated circulating water is return to the cooling tower for the heat removal. In this study, The design of Tower Cooling Water System is changed as follows : At First, The quantity of cells is changed into six in order to operate the cooling tower accurately correspond with condition of each equipment of head loads. The fans of cooling tower are controlled by the signal of TEW installed in the latter parts of it. The type of circulation water pump is modified to centrifugal pump and debris filter system is deleted.

Cooling of pressurized water nuclear reactor vessels

International Nuclear Information System (INIS)

Curet, H.D.

1978-01-01

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

Closed-cycle process of coke-cooling water in delayed coking unit

International Nuclear Information System (INIS)

Zhou, P.; Bai, Z.S.; Yang, Q.; Ma, J.; Wang, H.L.

2008-01-01

Synthesized processes are commonly used to treat coke-cooling wastewater. These include cold coke-cut water, diluting coke-cooling water, adding chemical deodorization into oily water, high-speed centrifugal separation, de-oiling and deodorization by coke adsorption, and open nature cooling. However, because of water and volatile evaporation loss, it is not suitable to process high-sulphur heavy oil using open treatments. This paper proposed a closed-cycling process in order to solve the wastewater treatment problem. The process is based on the characteristics of coke-cooling water, such as rapid parametric variation, oil-water-coke emulsification and steam-water mixing. The paper discussed the material characteristics and general idea of the study. The process of closed-cycle separation and utilization process of coke-cooling water was presented along with a process flow diagram. Several applications were presented, including a picture of hydrocyclones for pollution separation and a picture of equipments of pollution separation and components regeneration. The results showed good effect had been achieved since the coke-cooling water system was put into production in 2004. The recycling ratios for the components of the coke-cooling water were 100 per cent, and air quality in the operating area reached the requirements of the national operating site circumstance and the health standards. Calibration results of the demonstration unit were presented. It was concluded that since the devices went into operation, the function of production has been normal and stable. The operation was simple, flexible, adjustable and reliable, with significant economic efficiency and environmental benefits. 10 refs., 2 tabs., 3 figs

The potential for the recovery and reuse of cooling water in Taiwan

Energy Technology Data Exchange (ETDEWEB)

You, Shu-Hai; Tseng, Dyi-Hwa; Guo, Gia-Luen; Yang, Jyh-Jian [Graduate Institute of Environmental Engineering, National Central University, Chungli (Taiwan, Province of China)

1999-04-01

The cooling water is the major part of industrial water use in Taiwan, either from the view of demand priority or supply volume. In order to save water, the loading of supply system can be reduced if the cooling water can be recovered and reused. For this reason, exploration of the recent operation status of the cooling water system has become essential in Taiwan. This study was initially focused on the current applications and reuse trends of cooling water in oil refineries, chemical industry, steel mills, food industry, electronics works, textile plants and power stations. According to the statistical analysis, the portable water and groundwater are the primary sources of makeup water for cooling systems. The multiple-chemicals method and makeup treatment are increasingly accepted for the reclamation of cooling water. On the other hand, sidestream treatment and blowdown reuse are not popular in Taiwan. The recovery rate of blowdown is only 26.8%. The fact of higher cost is the major reason to depress the willingness of recovery. Some representative plants had been selected for case study. However, most cooling water systems are only operated by operator`s experience according to field investigation. In each case, the water quality indexes were used to evaluate the operational condition of cooling water systems. There was no case plant found to be operated at appropriate cycles of concentration. This paper also presented the bottlenecks of conservation technologies of cooling water in Taiwan. These bottlenecks include increasing the cycles of concentration, the reuse of wastewater, and the blowdown treatment for reuse. This paper also demonstrates that the recovery and reuse of cooling water has great potential and is feasible for the available technologies in present Taiwan, but the industries are still unwilling to upgrade because of initial cost. Finally, some approaches associated with technology, economics, environment and policy are proposed to be a

Experiences with the ASDEX neutralizer plates and construction of water-cooled plates for long-pulse heating

International Nuclear Information System (INIS)

Rapp, H.; Niedermeyer, H.; Kornherr, M.

1987-01-01

After dismantling of the titanium neutralizer plates inspection yielded satisfactory status of flat areas whereas edges and curved shapes were heavily melted. At the inner plates of the lower divertor strongly focused melting and cutting was found which is caused by fast electrons. These electrons are continuously produced. The production mechanism is not yet clear but runaway processes can be excluded. With long-pulse additional heating of 6 MW/10s as planned for ASDEX in 1987, the total energy delivered to the plasma will increase by a factor of 30. Therefore new water-cooled neutralizer plates have been constructed which consist of a copper-steel compound. The construction principle and the topology of the cooling circuits is presented

Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2016-01-01

The objective of this study is to develop a probabilistic risk assessment (PRA) methodology for extreme rainfall with focusing on decay heat removal system of a sodium-cooled fast reactor. For the extreme rainfall, annual excess probability depending on the hazard intensity was statistically estimated based on meteorological data. To identify core damage sequence, event trees were developed by assuming scenarios that structures, systems and components (SSCs) important to safety are flooded with rainwater coming into the buildings through gaps in the doors and the SSCs fail when the level of rainwater on the ground or on the roof of the building becomes higher than thresholds of doors on first floor or on the roof during the rainfall. To estimate the failure probability of the SSCs, the level of water rise was estimated by comparing the difference between precipitation and drainage capacity. By combining annual excess probability and the failure probability of SSCs, the event trees led to quantification of core damage frequency, and therefore the PRA methodology for rainfall was developed. (author)

MHD/gas turbine systems designed for low cooling water requirements

International Nuclear Information System (INIS)

Annen, K.D.; Eustis, R.H.

1983-01-01

The MHD/gas turbine combined-cycle system has been designed specifically for applications where the availability of cooling water is very limited. The base case systems which were studied consist of a coal-fired MHD plant with an air turbine bottoming plant and require no cooling water. In addition to the base case systems, systems were considered which included the addition of a vapor cycle bottoming plant to improve the thermal efficiency. These systems require a small amount of cooling water. The results show that the MHD/gas turbine systems have very good thermal and economic performances. The base case I MHD/gas turbine system (782 MW /SUB e/ ) requires no cooling water, has a heat rate which is 13% higher, and a cost of electricity which is only 7% higher than a comparable MHD/steam system (878 MW /SUB e/ ) having a cooling tower heat load of 720 MW. The case I vapor cycle bottomed systems have thermal and economic performances which approach and even exceed those of the MHD/steam system, while having substantially lower cooling water requirements. Performances of a second-generation MHD/gas turbine system and an oxygen-enriched, early commercial system are also evaluated. An analysis of nitric oxide emissions shows compliance with emission standards

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Modelling the radiolysis of RSG-GAS primary cooling water

Science.gov (United States)

Butarbutar, S. L.; Kusumastuti, R.; Subekti, M.; Sunaryo, G. R.

2018-02-01

Water chemistry control for light water coolant reactor required a reliable understanding of radiolysis effect in mitigating corrosion and degradation of reactor structure material. It is known that oxidator products can promote the corrosion, cracking and hydrogen pickup both in the core and in the associated piping components of the reactor. The objective of this work is to provide the radiolysis model of RSG GAS cooling water and further more to predict the oxidator concentration which can lead to corrosion of reactor material. Direct observations or measurements of the chemistry in and around the high-flux core region of a nuclear reactor are difficult due to the extreme conditions of high temperature, pressure, and mixed radiation fields. For this reason, chemical models and computer simulations of the radiolysis of water under these conditions are an important route of investigation. FACSIMILE were used to calculate the concentration of O2 formed at relatively long-time by the pure water γ and neutron irradiation (pH=7) at temperature between 25 and 50 °C. This simulation method is based on a complex chemical reaction kinetic. In this present work, 300 MeV-proton were used to mimic γ-rays radiolysis and 2 MeV fast neutrons. Concentration of O2 were calculated at 10-6 - 106 s time scale.

Cool-Water Carbonates, SEPM Special Publication No. 56

Science.gov (United States)

Hallock, Pamela

Doesn't field work on modern carbonates mean scuba diving on spectacular coral reefs in gin-clear water teeming with brightly colored fish? Not if you are one of the researchers that Jonathan Clarke of the Western Mining Corporation Ltd., in Preston, Victoria, Australia, assembled at a workshop in Geelong, Victoria, in January 1995. Their field work involves research cruises in high-latitude oceans, where mal de mer and chilling winds are constant companions. Many braved 10-m seas in modest-sized research vessels to sample shelves stripped of fine sediments by storm waves whose effects can reach to depths exceeding 200 m. Noel James of Queen's University in Kingston, Ontario, carefully lays the groundwork for the book in a paper titled, “The Cool-Water Carbonate Depositional Realm,” which will assuredly become a standard reading assignment in advanced undergraduate-and graduate-level courses in carbonate sedimentology. James skillfully shows how cool-water carbonates are part of the greater carbonate depositional spectrum. By expanding recognition of the possible range of carbonate environments, sedimentologists expand their ability to understand and interpret ancient carbonates, particularly Paleozoic limestones that often show striking similarities to modern cool-water sediments. James' paper is followed by nine papers on modern cool-water carbonates, seven on Tertiary environments, and seven examples from Mesozoic and Paleozoic limestones

Laboratory study on the cooling effect of flash water evaporative cooling technology for ventilation and air-conditioning of buildings

DEFF Research Database (Denmark)

Fang, Lei; Yuan, Shu; Yang, Jianrong

environments and the other simulated an air-conditioned indoor environment. The flash water evaporation cooling device was installed in the chamber that simulated indoor environment. The air from the chamber simulating outdoor environment was introduced into the cooling device and cooled by the flash water......, is effective for ventilation and air-conditioning in warm/hot and dry climate zones. The technology can provide fresh outdoor air with a temperature of 4 to 7 °C lower than room air temperature.......This paper presents a simple cooling technology using flash water evaporation. The technology combines a water atomizer with a plate heat exchanger used for heat recovery of a ventilation system. It is mainly used to cool the ventilation airflow from outdoors and is particularly suitable to be used...

The constructional design of cooling water discharge structures on German rivers

International Nuclear Information System (INIS)

Geldner, P.; Zimmermann, C.

1975-11-01

The present compilation of structures for discharging cooling water from power stations into rivers is an attempt to make evident developments in the constructional design of such structures and to give reasons for special structure shapes. A complete collection of all structures built in Germany, however, is difficult to realize because of the large number of power stations. For conventionally heated power stations therefore only a selection was made, while nuclear power stations in operation or under construction could almost completely be taken into account. For want of sufficient quantities of water for river water cooling, projected power stations are now almost exclusively designed for closed-circuit cooling so that the required discharge structures for elutrition water from the cooling towers as well as for the emergency and secondary cooling circuits have to be designed only for small amounts of water. (orig./HP) [de

WRI 50: Strategies for Cooling Electric Generating Facilities Utilizing Mine Water

Energy Technology Data Exchange (ETDEWEB)

Joseph J. Donovan; Brenden Duffy; Bruce R. Leavitt; James Stiles; Tamara Vandivort; Paul Ziemkiewicz

2004-11-01

Power generation and water consumption are inextricably linked. Because of this relationship DOE/NETL has funded a competitive research and development initiative to address this relationship. This report is part of that initiative and is in response to DOE/NETL solicitation DE-PS26-03NT41719-0. Thermal electric power generation requires large volumes of water to cool spent steam at the end of the turbine cycle. The required volumes are such that new plant siting is increasingly dependent on the availability of cooling circuit water. Even in the eastern U.S., large rivers such as the Monongahela may no longer be able to support additional, large power stations due to subscription of flow to existing plants, industrial, municipal and navigational requirements. Earlier studies conducted by West Virginia University (WV 132, WV 173 phase I, WV 173 Phase II, WV 173 Phase III, and WV 173 Phase IV in review) have identified that a large potential water resource resides in flooded, abandoned coal mines in the Pittsburgh Coal Basin, and likely elsewhere in the region and nation. This study evaluates the technical and economic potential of the Pittsburgh Coal Basin water source to supply new power plants with cooling water. Two approaches for supplying new power plants were evaluated. Type A employs mine water in conventional, evaporative cooling towers. Type B utilizes earth-coupled cooling with flooded underground mines as the principal heat sink for the power plant reject heat load. Existing mine discharges in the Pittsburgh Coal Basin were evaluated for flow and water quality. Based on this analysis, eight sites were identified where mine water could supply cooling water to a power plant. Three of these sites were employed for pre-engineering design and cost analysis of a Type A water supply system, including mine water collection, treatment, and delivery. This method was also applied to a ''base case'' river-source power plant, for comparison. Mine-water

Two-dimensional modeling of water spray cooling in superheated steam

Directory of Open Access Journals (Sweden)

Ebrahimian Vahid

2008-01-01

Full Text Available Spray cooling of the superheated steam occurs with the interaction of many complex physical processes, such as initial droplet formation, collision, coalescence, secondary break up, evaporation, turbulence generation, and modulation, as well as turbulent mixing, heat, mass and momentum transfer in a highly non-uniform two-phase environment. While it is extremely difficult to systematically study particular effects in this complex interaction in a well defined physical experiment, the interaction is well suited for numerical studies based on advanced detailed models of all the processes involved. This paper presents results of such a numerical experiment. Cooling of the superheated steam can be applied in order to decrease the temperature of superheated steam in power plants. By spraying the cooling water into the superheated steam, the temperature of the superheated steam can be controlled. In this work, water spray cooling was modeled to investigate the influences of the droplet size, injected velocity, the pressure and velocity of the superheated steam on the evaporation of the cooling water. The results show that by increasing the diameter of the droplets, the pressure and velocity of the superheated steam, the amount of evaporation of cooling water increases. .

Auxiliary equipment for cooling water in a reactor

International Nuclear Information System (INIS)

Konno, Yasuhiro; Sakairi, Toshiaki.

1975-01-01

Object: To effectively make use of pressure energy of reactor water, which has heretofore been discarded, to enable supply of emergency power supply of high reliability and to prevent spreading of environmental contamination. Structure: Sea water pumped by a sea water supply pump is fed to a heat exchanger. Reactor water carried through piping on the side to be cooled is removed in heat by the heat exchanger to be cooled and returned, and then again returned to the reactor. On the other hand, sea water heated by the heat exchanger is fed to a water wheel to drive the water wheel, after which it is discharged into a discharging path. A generator may be directly connected to the water wheel to use the electricity generated by the generator as the emergency power source. (Kamimura, M.)

40 CFR 63.1086 - How must I monitor for leaks to cooling water?

Science.gov (United States)

2010-07-01

... monitor for leaks to cooling water? You must monitor for leaks to cooling water by monitoring each heat... system so that the cooling water flow rate is 51,031 liters per minute or less so that a leak of 3.06 kg... detected a leak. (b) Individual heat exchangers. Monitor the cooling water at the entrance and exit of each...

An improved water cooled nuclear reactor and pressuriser assembly

International Nuclear Information System (INIS)

Gardner, F.J.; Strong, R.

1991-01-01

A water cooled nuclear reactor is described which comprises a reactor core, a primary water coolant circuit and a pressuriser arranged as an integral unit in a pressure vessel. The pressure vessel is divided into an upper and a lower chamber by a casing. The reactor core and primary water coolant circuit are arranged in the lower chamber and the pressuriser is arranged in the upper chamber. A plurality of spray pipes interconnect a steam space of the pressuriser with the downcomer of the primary water coolant circuit below a heat exchanger. A plurality of surge ports interconnect a water space of the pressuriser with the primary water coolant circuit. The surge ports have hydraulic diodes so that there is a low flow resistance for water from the water space of the pressuriser to the primary water coolant circuit and high flow resistance in the opposite direction. The spray pipes provide a desuperheating spray of cooled water into the pressuriser during positive volume surges of the primary water coolant. The pressuriser arrangement may also be applied to integral water cooled reactors with separate pressurisers and to dispersed pressurised water reactors. The surge ports also allow water to flow by gravity to the core in an emergency. (author)

Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor. Influence of flow collector of backup CR guide tube

International Nuclear Information System (INIS)

Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

2016-01-01

Design study of an advanced large-scale sodium-cooled fast reactor (SFR) has been conducted in JAEA. In the region between the bottom of the Upper Internal Structure (UIS) and the core outlet, the hot sodium from the fuel subassembly mixes with the cold sodium from the neighbor control rod (CR) channel. Therefore, temperature fluctuation due to mixing fluids at different temperatures may cause high cycle thermal fatigue at the bottom of the UIS. In the advanced design, installation of a flow guide structure named Flow-Collector (FC) to the backup control rod (BCR) guide tube is considered to enhance reliable operation of self-actuated shutdown system (SASS) and to ensure reactor shutdown operation. Previously, water experiments without the FC model had been examined in JAEA to investigate effective countermeasures to the significant temperature fluctuation generation at the bottom of the UIS. Since the FC may affect the thermal mixing behavior at the bottom of the UIS, influence of the FC on characteristics of the temperature fluctuation around the BCR channels was investigated using a water experimental facility with structure model of the FC. Through the experiment, small influence of the FC on the temperature fluctuation distribution at the bottom of the UIS was indicated. (author)

Process for cooling waste water

Energy Technology Data Exchange (ETDEWEB)

Rohner, P

1976-12-16

The process for avoiding thermal pollution of waters described rests on the principle of the heat conduction tube, by which heat is conducted from the liquid space into the atmosphere at a lower temperature above it. Such a tube, here called a cooling tube, consists in its simplest form of a heat conducting corrugated tube, made, for example, of copper or a copper alloy or of precious metals, which is sealed to be airtight at both ends, and after evacuation, is partially filled with a medium of low boiling point. The longer leg of the tube, which is bent at right angles, lies close below the surface of the water to be cooled and parallel to it; the shorter leg projects vertically into the atmosphere. The liquid inside the cooling tube fills the horizontal part of the tube to about halfway. A certain part of the liquid is always evaporated in this part. The vapor rising in the vertical part of the tube condenses on the internal wall cooled by the air outside, and gives off its heat to the atmosphere. The condensed medium flows back down the vertical internal wall into the initial position in a continuous cycle. A further development contains a smooth plastic inner tube in an outer corrugated tube, which is shorter than the outer tube; it ends at a distance from the caps sealing the outer tube at both ends. In this design the angle between the vertical and horizontal leg is less than 90/sup 0/. The shorter leg projects vertically from the water surface, below which the longer leg rises slightly from the knee of tube. The quantity of the liquid is gauged as a type of siphon, so that the space between the outer and inner tube at the knee of the tube remains closed by the liquid medium. The medium evaporated from the surface in the long leg of the tube therefore flows over the inner tube, which starts above the level of the medium. Thus evaporation and condensation paths are separated.

Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

Energy Technology Data Exchange (ETDEWEB)

Nagy, G. [KFKI Atomic Energy Research Institute, P.O. Box 49, Budapest H-1525 (Hungary)]. E-mail: nagyg@sunserv.kfki.hu; Somogyi, A. [KFKI Atomic Energy Research Institute, P.O. Box 49, Budapest H-1525 (Hungary); Patek, G. [Paks Nuclear Power Plant, P.O. Box 71, Paks H-7031 (Hungary); Pinter, T. [Paks Nuclear Power Plant, P.O. Box 71, Paks H-7031 (Hungary); Schiller, R. [KFKI Atomic Energy Research Institute, P.O. Box 49, Budapest H-1525 (Hungary)

2007-06-15

Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.

Preliminary Design of Compressor Impeller for innovative Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jekyoung; Cho, Seongkuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Cha, Jae Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

For nuclear power plant application, applying S-CO{sub 2} Brayton cycle to Sodium cooled Fast Reactors and Small Modular Reactors are currently considered and active research is being performed by various research institutions and universities. As a part of research activities on the SCO{sub 2} Brayton cycle development for a nuclear power system, KAIST joint research team is currently working on an innovative Sodium cooled Fast Reactor (iSFR) development which utilizes S-CO{sub 2} Brayton cycle as its power conversion system. Various research subjects including reactor physics, thermo-hydraulics, material, cycle analysis and system integration are being considered as research issues currently. However, technical issues rising from dramatic change of thermodynamic property of CO{sub 2} near the critical point still remain as problems to be solved. As a result, 3D impeller model generation based on 1D mean stream line analysis results was successfully performed for non-airfoil blades. Since 3D model generation module works successfully, KAIST{sub T}MD can support 3D CFD analysis for internal flow structure in the designed impeller. Compressor loss mechanisms are complex phenomena and these are difficulties to be modeled while considering each loss mechanism separately.

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

International Nuclear Information System (INIS)

Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; Lingerfelt, Eric J.; Wojtowicz, Anna

2015-01-01

Highlights: • Data analysis for high-performance simulations of reactors will be a problem that we address with a new management system. • We describe new input-output libraries for nuclear reactor simulations. • We describe a new user interface for visualizing and analyzing simulation results. • We show the utility of these systems with a 17 × 17 fuel assembly example simulation. • The availability of the code and avenues for collaboration are presented. - Abstract: Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system’s requirements and design

Operations improvement of the recycling water-cooling systems of sugar mills

Directory of Open Access Journals (Sweden)

Shcherbakov Vladimir Ivanovich

Full Text Available Water management in sugar factories doesn’t have analogues in its complexity among food industry enterprises. Water intensity of sugar production is very high. Circulation water, condensed water, pulp press water and others are used in technological processes. Water plays the main role in physical, chemical, thermotechnical processes of beet processing and sugar production. As a consequence of accession of Russia to the WTO the technical requirements for production processes are changing. The enforcements of ecological services to balance scheme of water consumption and water disposal increased. The reduction of fresh water expenditure is one of the main tasks in economy of sugar industry. The substantial role in fresh water expenditure is played by efficiency of cooling and aeration processes of conditionally clean waters of the 1st category. The article contains an observation of the technologies of the available solutions and recommendations for improving and upgrading the existing recycling water-cooling systems of sugar mills. The authors present the block diagram of the water sector of a sugar mill and a method of calculating the optimal constructive and technological parameters of cooling devices. Water cooling towers enhanced design and upgrades are offered.

An operational experience with cooling tower water system in chilling plant

International Nuclear Information System (INIS)

Rajan, Manju B.; Roy, Ankan; Ravi, K.V.

2015-01-01

Cooling towers are popular in industries as a very effective evaporative cooling technology for air conditioning. Supply of chilled water to air conditioning equipments of various plant buildings and cooling tower water to important equipments for heat removal is the purpose of chilling plant at PRPD. The cooling medium used is raw water available at site. Water chemistry is maintained by make-up and blowdown. In this paper, various observations made during plant operation and equipment maintenance are discussed. The issues observed was scaling and algal growth affecting the heat transfer and availability of the equipment. Corrosion related issues were observed to be less significant. Scaling indices were calculated to predict the behavior. (author)

Mathematical model and calculation of water-cooling efficiency in a film-filled cooling tower

Science.gov (United States)

Laptev, A. G.; Lapteva, E. A.

2016-10-01

Different approaches to simulation of momentum, mass, and energy transfer in packed beds are considered. The mathematical model of heat and mass transfer in a wetted packed bed for turbulent gas flow and laminar wave counter flow of the fluid film in sprinkler units of a water-cooling tower is presented. The packed bed is represented as the set of equivalent channels with correction to twisting. The idea put forward by P. Kapitsa on representation of waves on the interphase film surface as elements of the surface roughness in interaction with the gas flow is used. The temperature and moisture content profiles are found from the solution of differential equations of heat and mass transfer written for the equivalent channel with the volume heat and mass source. The equations for calculation of the average coefficients of heat emission and mass exchange in regular and irregular beds with different contact elements, as well as the expression for calculation of the average turbulent exchange coefficient are presented. The given formulas determine these coefficients for the known hydraulic resistance of the packed bed element. The results of solution of the system of equations are presented, and the water temperature profiles are shown for different sprinkler units in industrial water-cooling towers. The comparison with experimental data on thermal efficiency of the cooling tower is made; this allows one to determine the temperature of the cooled water at the output. The technical solutions on increasing the cooling tower performance by equalization of the air velocity profile at the input and creation of an additional phase contact region using irregular elements "Inzhekhim" are considered.

Forward osmosis applied to evaporative cooling make-up water

Energy Technology Data Exchange (ETDEWEB)

Nicoll, Peter; Thompson, Neil; Gray, Victoria [Modern Water plc, Guildford (United Kingdom)

2012-11-15

Modern Water is in the process of developing a number of forward osmosis based technologies, ranging from desalination to power generation. This paper outlines the progress made to date on the development and commercial deployment of a forward osmosis based process for the production of evaporative cooling tower make-up water from impaired water sources, including seawater. Evaporative cooling requires significant amounts of good quality water to replace the water lost by evaporation, drift and blowdown. This water can be provided by conventional desalination processes or by the use of tertiary treated sewage effluent. The conventional processes are well documented and understood in terms of operation and power consumption. A new process has been successfully developed and demonstrated that provides make-up water directly, using a core platform 'forward osmosis' technology. This new technology shows significant promise in allowing various raw water sources, such as seawater, to be used directly in the forward osmosis step, thus releasing the use of scarce and valuable high grade water for other more important uses. The paper presents theoretical and operational results for the process, where it is shown that the process can produce make-up water at a fraction of the operational expenditure when compared to conventional processes, in particular regarding power consumption, which in some cases may be as low as 15 % compared to competing processes. Chemical additives to the cooling water (osmotic agent) are retained within the process, thus reducing their overall consumption. Furthermore the chemistry of the cooling water does not support the growth of Legionella pneumophila. Corrosion results are also reported. (orig.)

Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

2017-04-20

The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

Review on Water Distribution of Cooling Tower in Power Station

Science.gov (United States)

Huichao, Zhang; Lei, Fang; Hao, Guang; Ying, Niu

2018-04-01

As the energy sources situation is becoming more and more severe, the importance of energy conservation and emissions reduction gets clearer. Since the optimization of water distribution system of cooling tower in power station can save a great amount of energy, the research of water distribution system gets more attention nowadays. This paper summarizes the development process of counter-flow type natural draft wet cooling tower and the water distribution system, and introduces the related domestic and international research situation. Combining the current situation, we come to the conclusion about the advantages and disadvantages of the several major water distribution modes, and analyze the problems of the existing water distribution ways in engineering application, furthermore, we put forward the direction of water distribution mode development on the basis knowledge of water distribution of cooling tower. Due to the water system can hardly be optimized again when it’s built, choosing an appropriate water distribution mode according to actual condition seems to be more significant.

Water cooled nuclear reactor

International Nuclear Information System (INIS)

1975-01-01

A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

Research and development of the supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Oka, Yoshiaki

2003-01-01

The concept of high temperature reactor cooled by light water (SCR) has been developed at the University of Tokyo since 1989. Major elements of reactor conceptual design and safety were studied. It includes fuel rod design, core design of thermal and fast reactors, plant heat balance, safety design, accident and transient analysis, LOCA, PSA, plant control, start-up and stability. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical FPP in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil Fired Power Plants (FPP) will be fully utilized for SCR. Although the concept was developed at the University of Tokyo mostly with our own funds and resources, four funding was/is provided for the research in Japan so far. Those are TEPCO studies with Japanese vendors in 1994 and 1995. JSPS (Monbusho) funding of pulse radiolysis of supercritical water to the University of Tokyo, Japanese-NERI program of METI to Toshiba team on thermal hydraulics, corrosion and plant system and Japanese-NERI program of MEXT on water chemistry to the University of Tokyo. The concept was taken as the reference of HPLWR study in Europe with funding of EU in 2000 and 2001. The concept was evaluated in the Generation 4 reactor program in USA. It was selected as only one water-cooled Generation 4 reactor. This paper describes the overview of the conceptual design at the University of Tokyo and R and D in the world

Minor actinide burning in dedicated lead-bismuth cooled fast reactors

International Nuclear Information System (INIS)

Hejzlar, P.; Driscoll, M.J.; Kazimi, M.S.; Todreas, N.E.

2001-01-01

The destruction of minor actinides (MA) in dedicated burners is of contemporary interest in Europe and Japan because it requires the deployment of smaller number of special transmutation facilities. A major fraction of Pu from spent LWR fuel can be then burned in PWRs (or fast reactors) using dedicated fertile-free fuel assemblies. However, the design of MA burning fast spectrum cores poses significant challenges because of deterioration of key safety parameters, in particular of the coolant void coefficient. This study proposes the concept of an lead-bismuth eutectic (LBE)-cooled dedicated MA burner having metallic fuel (MA-Pu-Zr) and streaming assemblies to attain acceptable coolant void worth performance. It is shown that a large 1800 MWth fertile-free core containing 37 wt% TRU with very high fraction of MA(59 wt%) from LWR spent fuel can be burned in a first cycle for 700 EFPDs with a very small reactivity swing: less than β eff . Moreover, the reactivity void worth is negative for a fully voided core when all surrounding coolant is kept at reference density. However, the core reactivity increases as coolant density falls from the reference value of 10.25 to 6 g/cm 3 . Because its coolant density coefficient value is less than that of a sodium cooled IFR, the concept provides good potential for the achievement of self-regulation characteristics in unprotected events, provided that small negative fuel temperature feedback can be maintained. (authors)

Cooling water injection system

International Nuclear Information System (INIS)

Inai, Nobuhiko.

1989-01-01

In a BWR type reactor, ECCS system is constituted as a so-called stand-by system which is not used during usual operation and there is a significant discontinuity in relation with the usual system. It is extremely important that ECCS operates upon occurrence of accidents just as specified. In view of the above in the present invention, the stand-by system is disposed along the same line with the usual system. That is, a driving water supply pump for supplying driving water to a jet pump is driven by a driving mechanism. The driving mechanism drives continuously the driving water supply pump in a case if an expected accident such as loss of the function of the water supply pump, as well as during normal operation. That is, all of the water supply pump, jet pump, driving water supply pump and driving mechanism therefor are caused to operate also during normal operation. The operation of them are not initiated upon accident. Thus, the cooling water injection system can perform at high reliability to remarkably improve the plant safety. (K.M.)

Water spray cooling technique applied on a photovoltaic panel: The performance response

International Nuclear Information System (INIS)

Nižetić, S.; Čoko, D.; Yadav, A.; Grubišić-Čabo, F.

2016-01-01

Highlights: • An experimental study was conducted on a monocrystalline photovoltaic panel (PV). • A water spray cooling technique was implemented to determine PV panel response. • The experimental results showed favorable cooling effect on the panel performance. • A feasibility aspect of the water spray cooling technique was also proven. - Abstract: This paper presents an alternative cooling technique for photovoltaic (PV) panels that includes a water spray application over panel surfaces. An alternative cooling technique in the sense that both sides of the PV panel were cooled simultaneously, to investigate the total water spray cooling effect on the PV panel performance in circumstances of peak solar irradiation levels. A specific experimental setup was elaborated in detail and the developed cooling system for the PV panel was tested in a geographical location with a typical Mediterranean climate. The experimental result shows that it is possible to achieve a maximal total increase of 16.3% (effective 7.7%) in electric power output and a total increase of 14.1% (effective 5.9%) in PV panel electrical efficiency by using the proposed cooling technique in circumstances of peak solar irradiation. Furthermore, it was also possible to decrease panel temperature from an average 54 °C (non-cooled PV panel) to 24 °C in the case of simultaneous front and backside PV panel cooling. Economic feasibility was also determined for of the proposed water spray cooling technique, where the main advantage of the analyzed cooling technique is regarding the PV panel’s surface and its self-cleaning effect, which additionally acts as a booster to the average delivered electricity.

Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

International Nuclear Information System (INIS)

Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

1996-01-01

The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

Design of a spherical fuel element for a gas-cooled fast reactor

International Nuclear Information System (INIS)

Van Rooijen, W.F.G.; Kloosterman, J.L.; Van Dam, H.; Van der Hagen, T.H.J.J.

2004-01-01

A study is undertaken to develop a fuel cycle for a gas-cooled fast reactor (GCFR). The design goals are: highly efficient use of (depleted) uranium, application of Pu recycled from LWR discharge as fissile material, high temperature output and simplicity of design. The design focuses on spherical TRISO-like fuel elements, a homogeneous core at start-up, providing for easy fuel fabrication, and self-breeding capability with a flat k eff with burn-up. Nitride fuel ( 15 N > 99%) has been selected because of its favourable thermal conductivity, high heavy metal density and compatibility with PUREX reprocessing. Two core concepts have been studied: one with coated particles embedded inside fuel pebbles, and one with coated particles cooled directly by helium. The result is that a flat k eff can be achieved for a long period of time, using coated particles cooled directly, with a homogeneous core at, start-up, with a closed fuel cycle and a simple refuelling and reprocessing scheme. (author)

Evaluation of a Design Concept for the Combined Air-water Passive Cooling PAFS+

International Nuclear Information System (INIS)

Bae, Sung Won; Kwon, Taesoon

2014-01-01

The APR+ system provides the Passive Auxiliary Feed-water System (PAFS) for the passive cooling capability. However, the current design requirement for working time for the PAFS is about 8 hours only. Thus, current working time of PAFS can not meet the required 72 hours cooling capability for the long term SBO situation. To meet the 72 hours cooling, the pool capacity should be almost 3∼4 times larger than that of current water cooling tank. In order to continue the PAFS operation for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility study on the combined passive air-water cooling system. Figure 1 and 2 show the conceptual difference of the PAFS and combined passive air-water cooling system, respectively. Simple performance evaluation of the passive air cooling heat exchanger has been conducted by the MARS calculation. For the postulated FLB scenario, 4800 heat exchanger tubes and 5 m/s air velocity are not sufficient to sustain the PCCT pool level for 72 hour cooling. Further works on the system design and performance enhancing plan are required to fulfill the 72 hours long term passive cooling

Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

International Nuclear Information System (INIS)

Lap-Yan, C.; Wie, T. Y. C.

2009-01-01

The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs) is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR) in a GEN IV direct-cycle gas-cooled fast reactor (GFR) which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

The effects of high-Ca hardness water treatment for secondary cooling water in HANARO

International Nuclear Information System (INIS)

Kang, T. J.; Park, Y. C.; Hwang, S. R.; Lim, I. C.; Choi, H. Y.

2003-01-01

Water-quality control of the second cooling system in HANARO has been altered from low Ca-hardness treatment to high Ca-hardness treatment since March, 2001. High Ca-hardness water treatment in HANARO is to maintain the calcium hardness around 12 by minimizing the blowdown of secondary cooling water. This paper describes the effect of cost reduction after change of water-quility treatment method. The result shows that the cost of the water could be reduced by 25% using the pond water in KAERI. The amount and cost for the chemical agent could be reduced by 40% and 10% respectively

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Radionuclides behaviour in the silts-water system of a cooling pond

International Nuclear Information System (INIS)

Ol'khovik, Yu.A.; Kostyuchenko, N.G.; Koromyslichenko, T.I.

1989-01-01

As a result of the Chernobyl' accident a considerable amount of radioisotopes (1-5x10 5 Ci) concentrated in a cooling pond. A year later the accident a level of water contamination decreased by 2 orders, whereas the radionuclide distribution changed perceptibly. Processes of water self-decontamination in the cooling pond were considered. A forecast of water radiactivity level in the cooling pond in the summer of 1988 was made. 3 refs.; 1 refs.; 2 tabs

Pavement-Watering for Cooling the Built Environment: A Review

OpenAIRE

Hendel , Martin

2016-01-01

Pavement-watering is being considered by decision-makers in many cities as a means of cooling the built environment and of adapting to rising extreme heat events due to climate change. In this article we review the existing literature on the topic of pavement-watering. We first focus on the methodological choices made in the literature, including study approach and scale, watering methods used as well as how results are analyzed. We then discuss the cooling effects reported, separating micro-...

Water conservation and improved production efficiency using closed-loop evaporative cooling systems

Energy Technology Data Exchange (ETDEWEB)

Marchetta, C. [Niagara Blower Co., Buffalo, NY (United States)

2009-07-01

This paper described wet surface air coolers (WSAC) that can be used in refineries and hydrocarbon processing plants to address water use issues. These closed-loop evaporative cooling systems are a cost-effective technology for both heat transfer and water conservation. WSACs can help deliver required cooling water temperatures and improve plant performance while using water streams currently considered to be unusable with conventional towers and heat exchangers. WSACs are versatile and can provide solutions to water use, water quality, and outlet temperature. The benefits of the WSAC include capital cost savings, reduced system pressures, lower carbon footprint, and the ability to use poor quality water as makeup. Water makeup can be blowdown from other equipment, plant effluent, reclaimed water, produced water, flue gas desulphurization (FGD) wastewater, and even seawater. Units can be manufactured with a wide variety of materials depending on water quality, water treatment, and cycles of concentration. This paper also provided comparisons to other alternative technologies, capital and operating cost savings, and site specific case studies. Two other system designs can accommodate closed-loop heat transfer applications, notably an open tower with a heat exchanger and a dry, air-cooled system. A WSAC system is an efficient and effective heat rejection technology for several reasons. The WSAC cooler or condenser utilizes latent cooling, which is far more efficient than sensible cooling. This means that a WSAC system can cool the same heat load with a smaller footprint than all-dry systems. 6 figs.

CALIOP: a multichannel design code for gas-cooled fast reactors. Code description and user's guide

International Nuclear Information System (INIS)

Thompson, W.I.

1980-10-01

CALIOP is a design code for fluid-cooled reactors composed of parallel fuel tubes in hexagonal or cylindrical ducts. It may be used with gaseous or liquid coolants. It has been used chiefly for design of a helium-cooled fast breeder reactor and has built-in cross section information to permit calculations of fuel loading, breeding ratio, and doubling time. Optional cross-section input allows the code to be used with moderated cores and with other fuels

Cooling tower make-up water processing for nuclear power plants: a comparison

Energy Technology Data Exchange (ETDEWEB)

Andres, O; Flunkert, F; Hampel, G; Schiffers, A [Rheinisch-Westfaelisches Elektrizitaetswerk A.G., Essen (Germany, F.R.)

1977-01-01

In water-cooled nuclear power plants, 1 to 2% of the total investment costs go to cooling tower make-up water processing. The crude water taken from rivers or stationary waters for cooling must be sufficiently purified regarding its content of solids, carbonate hardness and corrosive components so as to guarantee an operation free of disturbances. At the same time, the processing methods must be selected for operational-economic reasons in such a manner that waste water and waste problems are kept small regarding environmental protection. The various parameters described have a decisive influence on the processing methods of the crude water, individual processes (filtration, sedimentation, decarbonization) are described, circuit possibilities for cooling water systems are compared and the various processes are analyzed and compared with regard to profitableness and environmental compatability.

Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors

International Nuclear Information System (INIS)

Sackett, J.I.

1983-01-01

This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance

Investigation on flow stability of supercritical water cooled systems

International Nuclear Information System (INIS)

Cheng, X.; Kuang, B.

2006-01-01

Research activities are ongoing worldwide to develop nuclear power plants with supercritical water cooled reactor (SCWR) with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, the strong variation of the thermal-physical properties of water in the vicinity of the pseudo-critical line results in challenging tasks in various fields, e.g. thermal-hydraulic design of a SCWR. One of the challenging tasks is to understand and to predict the dynamic behavior of supercritical water cooled systems. Although many thermal-hydraulic research activities were carried out worldwide in the past as well as in the near present, studies on dynamic behavior and flow stability of SC water cooled systems are scare. Due to the strong density variation, flow stability is expected to be one of the key items which need to be taken into account in the design of a SCWR. In the present work, the dynamic behavior and flow stability of SC water cooled systems are investigated using both numerical and theoretical approaches. For this purpose a new computer code SASC was developed, which can be applied to analysis the dynamic behavior of systems cooled by supercritical fluids. In addition, based on the assumptions of a simplified system, a theoretical model was derived for the prediction of the onset of flow instability. A comparison was made between the results obtained using the theoretical model and those from the SASC code. A good agreement was achieved. This gives the first evidence of the reliability of both the SASC code and the theoretical model

Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

International Nuclear Information System (INIS)

Lechelle, J.; Aufore, L.; Basini, V.; Belin, R.; Vaudez, S.

2004-01-01

Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

International Nuclear Information System (INIS)

Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

2016-01-01

Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

Gas cooled fast reactor benchmarks for JNC and Cea neutronic tools assessment

International Nuclear Information System (INIS)

Rimpault, G.; Sugino, K.; Hayashi, H.

2005-01-01

In order to verify the adequacy of JNC and Cea computational tools for the definition of GCFR (gas cooled fast reactor) core characteristics, GCFR neutronic benchmarks have been performed. The benchmarks have been carried out on two different cores: 1) a conventional Gas-Cooled fast Reactor (EGCR) core with pin-type fuel, and 2) an innovative He-cooled Coated-Particle Fuel (CPF) core. Core characteristics being studied include: -) Criticality (Effective multiplication factor or K-effective), -) Instantaneous breeding gain (BG), -) Core Doppler effect, and -) Coolant depressurization reactivity. K-effective and coolant depressurization reactivity at EOEC (End Of Equilibrium Cycle) state were calculated since these values are the most critical characteristics in the core design. In order to check the influence due to the difference of depletion calculation systems, a simple depletion calculation benchmark was performed. Values such as: -) burnup reactivity loss, -) mass balance of heavy metals and fission products (FP) were calculated. Results of the core design characteristics calculated by both JNC and Cea sides agree quite satisfactorily in terms of core conceptual design study. Potential features for improving the GCFR computational tools have been discovered during the course of this benchmark such as the way to calculate accurately the breeding gain. Different ways to improve the accuracy of the calculations have also been identified. In particular, investigation on nuclear data for steel is important for EGCR and for lumped fission products in both cores. The outcome of this benchmark is already satisfactory and will help to design more precisely GCFR cores. (authors)

Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

2013-03-01

'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

Simulation study of air and water cooled photovoltaic panel using ANSYS

Science.gov (United States)

Syafiqah, Z.; Amin, N. A. M.; Irwan, Y. M.; Majid, M. S. A.; Aziz, N. A.

2017-10-01

Demand for alternative energy is growing due to decrease of fossil fuels sources. One of the promising and popular renewable energy technology is a photovoltaic (PV) technology. During the actual operation of PV cells, only around 15% of solar irradiance is converted to electricity, while the rest is converted into heat. The electrical efficiency decreases with the increment in PV panel’s temperature. This electrical energy is referring to the open-circuit voltage (Voc), short-circuit current (Isc) and output power generate. This paper examines and discusses the PV panel with water and air cooling system. The air cooling system was installed at the back of PV panel while water cooling system at front surface. The analyses of both cooling systems were done by using ANSYS CFX and PSPICE software. The highest temperature of PV panel without cooling system is 66.3 °C. There is a decrement of 19.2% and 53.2% in temperature with the air and water cooling system applied to PV panel.

Cooling tower water conditioning study. [using ozone

Science.gov (United States)

Humphrey, M. F.; French, K. R.

1979-01-01

Successful elimination of cooling tower treatment chemicals was demonstrated. Three towers functioned for long periods of time with ozone as the only treatment for the water. The water in the systems was reused as much as 30 times (cycles of concentration) without deleterious effects to the heat exchangers. Actual system blow-down was eliminated and the only makeup water added was that required to replace the evaporation and mist entrainment losses. Minimum water savings alone are approximately 75.1 1/kg/year. Cost estimates indicate that a savings of 55 percent was obtained on the systems using ozone. A major problem experienced in the use of ozone for cooling tower applications was the difficulty of accurate concentration measurements. The ability to control the operational characteristics relies on easily and accurately determined concentration levels. Present methods of detection are subject to inaccuracies because of interfering materials and the rapid destruction of the ozone.

Increasing photovoltaic panel power through water cooling technique

Directory of Open Access Journals (Sweden)

Calebe Abrenhosa Matias

2017-02-01

Full Text Available This paper presents the development of a cooling apparatus using water in a commercial photovoltaic panel in order to analyze the increased efficiency through decreased operating temperature. The system enables the application of reuse water flow, at ambient temperature, on the front surface of PV panel and is composed of an inclined plane support, a perforated aluminum profile and a water gutter. A luminaire was specially developed to simulate the solar radiation over the module under test in a closed room, free from the influence of external climatic conditions, to carry out the repetition of the experiment in controlled situations. The panel was submitted to different rates of water flow. The best water flow rate was of 0.6 L/min and net energy of 77.41Wh. Gain of 22.69% compared to the panel without the cooling system.

Reduction of Langelier index of cooling water by electrolytic ...

African Journals Online (AJOL)

LSI) of the cooling water from a cooling tower of a textile industry was investigated. Sacrificial anodes were employed which prevent obnoxious chlorine generation. A series of batch experiments using stainless steel electrodes were conducted ...

Water conservation benefits of urban heat mitigation: can cooling strategies reduce water consumption in California?

Science.gov (United States)

Vahmani, P.; Jones, A. D.

2017-12-01

Urban areas are at the forefront of climate mitigation and adaptation efforts given their high concentration of people, industry, and infrastructure. Many cities globally are seeking strategies to counter the consequences of both a hotter and drier climate. While urban heat mitigation strategies have been shown to have beneficial effects on health, energy consumption, and greenhouse gas emissions, their implications for water conservation have not been widely examined. Here we show that broad implementation of cool roofs, an urban heat mitigation strategy, not only results in significant cooling of air temperature, but also meaningfully decreases outdoor water consumption by reducing evaporative and irrigation water demands. Based on a suite of satellite-supported, multiyear regional climate simulations, we find that cool roof adoption has the potential to reduce outdoor water consumption across the major metropolitan areas in California by up to 9%. Irrigation water savings per capita, induced by cool roofs, range from 1.8 to 15.4 gallons per day across 18 counties examined. Total water savings in Los Angeles county alone is about 83 million gallons per day. While this effect is robust across the 15 years examined (2001-2015), including both drought and non-drought years, we find that cool roofs are most effective during the hottest days of the year, indicating that they could play an even greater role in reducing outdoor water use in a hotter future climate. We further show that this synergistic relationship between heat mitigation and water conservation is asymmetrical - policies that encourage direct reductions in irrigation water use can lead to substantial regional warming, potentially conflicting with heat mitigation efforts designed to counter the effects of the projected warming climate.

Cooling performance of helium-gas/water coolers in HENDEL

International Nuclear Information System (INIS)

Inagaki, Yoshiyuki; Takada, Shoji; Hayashi, Haruyoshi; Kobayashi, Toshiaki; Ohta, Yukimaru; Shimomura, Hiroaki; Miyamoto, Yoshiaki

1994-01-01

The helium engineering demonstration loop (HENDEL) has four helium-gas/water coolers where the cooling water flows in the tubes and helium gas on the shell side. Their cooling performance was studied using the operational data from 1982 to 1991. The heat transfer of helium gas on the shell was obtained for segmental and step-up baffle type coolers. Also, the change with operation time was investigated. The cooling performance was lowered by the graphite powder released from the graphite components for several thousand hours and thereafter recovered because the graphite powder from the components was reduced and the powder in the cooler shell was blown off during the operation. (orig.)

Creating prototypes for cooling urban water bodies

NARCIS (Netherlands)

Cortesoao, Joao; Klok, E.J.; Lenzholzer, Sanda; Jacobs, C.M.J.; Kluck, J.

2017-01-01

Abstract When addressing urban heat problems, climate- conscious urban design has been assuming that urban water bodies such as canals, ditches or ponds cool down their surroundings. Recent research shows that this is not necessarily the case and that urban water bodies may actually have a warming e

Contrastive analysis of cooling performance between a high-level water collecting cooling tower and a typical cooling tower

Science.gov (United States)

Wang, Miao; Wang, Jin; Wang, Jiajin; Shi, Cheng

2018-02-01

A three-dimensional (3D) numerical model is established and validated for cooling performance optimization between a high-level water collecting natural draft wet cooling tower (HNDWCT) and a usual natural draft wet cooling tower (UNDWCT) under the actual operation condition at Wanzhou power plant, Chongqing, China. User defined functions (UDFs) of source terms are composed and loaded into the spray, fill and rain zones. Considering the conditions of impact on three kinds of corrugated fills (Double-oblique wave, Two-way wave and S wave) and four kinds of fill height (1.25 m, 1.5 m, 1.75 m and 2 m), numerical simulation of cooling performance are analysed. The results demonstrate that the S wave has the highest cooling efficiency in three fills for both towers, indicating that fill characteristics are crucial to cooling performance. Moreover, the cooling performance of the HNDWCT is far superior to that of the UNDWCT with fill height increases of 1.75 m and above, because the air mass flow rate in the fill zone of the HNDWCT improves more than that in the UNDWCT, as a result of the rain zone resistance declining sharply for the HNDWCT. In addition, the mass and heat transfer capacity of the HNDWCT is better in the tower centre zone than in the outer zone near the tower wall under a uniform fill layout. This behaviour is inverted for the UNDWCT, perhaps because the high-level collection devices play the role of flow guiding in the inner zone. Therefore, when non-uniform fill layout optimization is applied to the HNDWCT, the inner zone increases in height from 1.75 m to 2 m, the outer zone reduces in height from 1.75 m to 1.5 m, and the outlet water temperature declines approximately 0.4 K compared to that of the uniform layout.

Cooling water treatment for heavy water project (Paper No. 6.9)

International Nuclear Information System (INIS)

Valsangkar, H.N.

1992-01-01

With minor exceptions, water is the preferred industrial medium for the removal of unwanted heat from process systems. The application of various chemical treatments is required to protect the system from water related and process related problems of corrosion, scale and deposition and biofouling. The paper discusses the cooling water problems for heavy water industries along with the impact caused by associated fertilizer units. (author). 6 figs

Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

International Nuclear Information System (INIS)

Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira; Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

2015-01-01

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Graduate School of Engineering, The University of Tokyo, Ibaraki (Japan); Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki [Japan Atomic Energy Agency, Ibaraki (Japan)

2015-10-15

Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

International Nuclear Information System (INIS)

Philpott, E.F.; Pounder, F.; Willby, C.R.

1978-01-01

The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

Science.gov (United States)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Conceptual design study on simplified and safer cooling systems for sodium cooled FBRs

International Nuclear Information System (INIS)

Hayafune, Hiroki; Shimakawa, Yoshio; Ishikawa, Hiroyasu; Kubota, Kenichi; Kobayashi, Jun; Kasai, Shigeo

2000-06-01

The objective of this study is to create the FBR plant concepts increasing economy and safety for the Phase-I 'Feasibility Studies on Commercialized Fast Reactor System'. In this study, various concepts of simplified 2ry cooling system for sodium cooled FBRs are considered and evaluated from the view points of technological feasibility, economy, and safety. The concepts in the study are considered on the basis of the following points of view. 1. To simplify 2ry cooling system by moderating and localizing the sodium-water reaction in the steam generator of the FBRs. 2. To simplify 2ry cooling system by eliminating the sodium-water reaction using integrated IHX-SG unit. 3. To simplify 2ry cooling system by eliminating the sodium-water reaction using a power generating system other than the steam generator. As the result of the study, 12 concepts and 3 innovative concepts are proposed. The evaluation study for those concepts shows the following technical prospects. 1. 2 concepts of integrated IHX-SG unit can eliminate the sodium-water reaction. Separated IHX and SG tubes unit using Lead-Bismuth as the heat transfer medium. Integrated IHX-SG unit using copper as the heat transfer medium. 2. Cost reduction effect by simplified 2ry cooling system using integrated IHX-SG unit is estimated 0 to 5%. 3. All of the integrated IHX-SG unit concepts have more weight and larger size than conventional steam generator unit. The weight of the unit during transporting and lifting would limit capacity of heat transfer system. These evaluation results will be compared with the results in JFY 2000 and used for the Phase-II study. (author)

IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

International Nuclear Information System (INIS)

2010-01-01

The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

Development of gas-cooled fast reactor and its thermo-hydraulics

International Nuclear Information System (INIS)

Kawamura, Hiroshi

1977-10-01

Development, thermo-hydraulics and safety of GCFR are reviewed. The Development of Gas-Cooled Fast Reactor (GCFR) utilizes helium technology of HTGR and fuel technology of LMFBR. The breeding ratio of GCFR will be larger than that of LMFBR by about 0.2. Features of GCFR are a fuel with roughened surface to raise the heat transfer and vent system for the pressure equalization in the fuel rod. Helium as coolant of GCFR is chemically stable and stays in the single phase. So, there is no fuel-coolant interaction unlike the case of LMFBR. Since the helium must be pressurized, possibility of a depressurization accident is not negligible. In the United States, a 300MWe demonstration plant program is about to start; the collaboration with European countries is now quite active in this field. Though the development of GCFR started behind that of LMFBR, GCFR is equally promising as a fast breeder reactor. When realized, it will present possibility of a choice between these two. (auth.)

Distinct difference of flaA genotypes of Legionella pneumophila between isolates from bath water and cooling tower water.

Science.gov (United States)

Amemura-Maekawa, Junko; Kura, Fumiaki; Chang, Bin; Suzuki-Hashimoto, Atsuko; Ichinose, Masayuki; Endo, Takuro; Watanabe, Haruo

2008-09-01

To investigate the genetic difference of Legionella pneumophila in human-made environments, we collected isolates of L. pneumophila from bath water (n = 167) and cooling tower water (n = 128) primarily in the Kanto region in 2001 and 2005. The environmental isolates were serogrouped and sequenced for a target region of flaA. A total of 14 types of flaA genotypes were found: 10 from cooling tower water and nine from bath water. The flaA genotypes of isolates from cooling tower water were quite different from those of bath water.

District cool water distribution; Reseau urbain et distribution d`eau glacee

Energy Technology Data Exchange (ETDEWEB)

Schabaillie, D. [Ste Climespace (France)

1997-12-31

The city of Paris has developed several district cool water distribution networks (Climespace) for air conditioning purposes, one in the Halles district (central Paris) linked with the Louvre museum, one in the Opera district (with large department stores) and one in the east of paris (Bercy). Each of these networks has a cool water production plant, the one at the Halles producing also hot water and safety electric power. The characteristics of the equipment (heat pumps, refrigerating machinery, storage...) are described. The pipes are laid in the city sewage network, and the cool carrier is water. The various networks are centrally supervised at the Halles center

Gas cooled fast reactor background, facilities, industries and programmes

International Nuclear Information System (INIS)

Dalle Donne, M.

1980-05-01

This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.) [de

Numerical study of coupled heat and mass transfer in geothermal water cooling tower

International Nuclear Information System (INIS)

Bourouni, K.; Bassem, M.M.; Chaibi, M.T.

2008-01-01

Cross flow mechanical cooling towers, widely spreads all over the south region of Tunisia are used for cooling geothermal water for agriculture and domestic ends. These towers are sized empirically and present several problems in regard to operation and electrical energy consumption. This work aims to study the thermal behaviour of this type of cooling towers through a developed mathematical model considering the variation of the water mass flow rate inside the tower. The analysis of the water and air temperatures distribution along the cooling tower had underlined the negative convection phenomenon at a certain height of the tower. This analysis has shown also that the difference in water temperature between the inlet and the outlet of the tower is much higher than the one of air due to the dominance of the evaporative potential compared to the convective one. In addition, the variations of the air humidity along the cooling tower and the quantity of evaporated water have been investigated. The loss of water by evaporation is found to be 5.1% of the total quantity of water feeding the cooling tower. Interesting future prospects are expected for validation of the developed model to optimize the operating of the cooling tower

Supplementary report: cooling water systems for Darlington G.S

International Nuclear Information System (INIS)

1975-08-01

This report summarizes Ontario Hydro's existing aquatic environmental programs, presents results of these investigations, and outlines plans and activities for expanded aquatic environment studies including the evaluation of alternative cooling systems. This report outlines specific considerations regarding possible alternative cooling arrangements for the Darlington station. It concludes with a recommendation that a study be initiated to examine the potential benefits of using the heated discharge water in a warm water recreational centre. (author)

MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core

International Nuclear Information System (INIS)

Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin

2005-01-01

Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to

Evaluation of heat exchange performance for the auxiliary component cooling water system cooling tower in HTTR

International Nuclear Information System (INIS)

Tochio, Daisuke; Kameyama, Yasuhiko; Shimizu, Atsushi; Inoi, Hiroyuki; Yamazaki, Kazunori; Shimizu, Yasunori; Aragaki, Etsushi; Ota, Yukimaru; Fujimoto, Nozomu

2006-09-01

The auxiliary component cooling water system (ACCWS) is one of the cooling system in High Temperature Engineering Test Reactor (HTTR). The ACCWS has main two features, many facilities cooling, and heat sink of the vessel cooling system which is one of the engineering safety features. Therefore, the ACCWS is required to satisfy the design criteria of heat removal performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the ACCWS cooling tower was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that ACCWS cooling tower has the required heat exchange performance in the design. (author)

The determinants of thermal comfort in cool water.

Science.gov (United States)

Guéritée, J; House, J R; Redortier, B; Tipton, M J

2015-10-01

Water-based activities may result in the loss of thermal comfort (TC). We hypothesized that in cooling water, the hands and feet would be responsible. Supine immersions were conducted in up to five clothing conditions (exposing various regions), as well as investigations to determine if a "reference" skin temperature (Tsk) distribution in thermoneutral air would help interpret our findings. After 10 min in 34.5 °C water, the temperature was decreased to 19.5 °C over 20 min; eight resting or exercising volunteers reported when they no longer felt comfortable and which region was responsible. TC, rectal temperature, and Tsk were measured. Rather than the extremities, the lower back and chest caused the loss of overall TC. At this point, mean (SD) chest Tsk was 3.3 (1.7) °C lower than the reference temperature (P = 0.005), and 3.8 (1.5) °C lower for the back (P = 0.002). Finger Tsk was 3.1 (2.7) °C higher than the reference temperature (P = 0.037). In cool and cooling water, hands and feet, already adapted to colder air temperatures, will not cause discomfort. Contrarily, more discomfort may arise from the chest and lower back, as these regions cool by more than normal. Thus, Tsk distribution in thermoneutral air may help understand variations in TC responses across the body. © 2014 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

Fluid Induced Vibration Analysis of a Cooling Water Pipeline for the HANARO CNS

International Nuclear Information System (INIS)

Kim, Bong Soo; Lee, Young Sub; Kim, Ik Soo; Kim, Young Ki

2007-01-01

CNS is the initial of Cold Neutron Source and the CNS facility system consists of hydrogen, a vacuum, a gas blanketing, a helium refrigeration and a cooling water supply system. Out of these subsystems, the helium refrigeration system has the function of removal of heat from a thermal neutron under reactor operation. Therefore, HRS (helium refrigeration system) must be under normal operation for the production of cold neutron. HRS is mainly made up of a helium compressor and a coldbox. This equipment is in need of cooling water to get rid of heat generation under stable operation and a cooling water system is essential to maintain the normal operation of a helium compressor and a coldbox. The main problem for the cooling water system is the vibration issue in the middle of operation due to a water flow in a pipeline. In order to suppress the vibration problem for a pipeline, the characteristics of a pipeline and fluid flow must be analyzed in detail. In this paper, fluid induced vibration of a cooling water pipe is analyzed numerically and the stability of the cooling water pipeline is investigated by using pipe dynamic theory

Detailed Design of Cooling Water System for Cold Neutron Source in HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, Bong Soo; Choi, Jung Woon; Kim, Y. K.; Wu, S. I.; Lee, Y. S

2007-04-15

To make cold neutron, a cryogenic refrigerator is necessary to transform moderator into cryogenic state so, thermal neutron is changed into cold neutron through heat transfer with moderator. A cryogenic refrigerator mainly consists of two apparatus, a helium compressor and a cold box which needs supply of cooling water. Therefore, cooling water system is essential to operate of cryogenic refrigerator normally. This report is mainly focused on the detailed design of the cooling water system for the HANARO cold neutron source, and describes design requirement, calculation, specification of equipment and water treatment method.

Detailed Design of Cooling Water System for Cold Neutron Source in HANARO

International Nuclear Information System (INIS)

Kim, Bong Soo; Choi, Jung Woon; Kim, Y. K.; Wu, S. I.; Lee, Y. S.

2007-04-01

To make cold neutron, a cryogenic refrigerator is necessary to transform moderator into cryogenic state so, thermal neutron is changed into cold neutron through heat transfer with moderator. A cryogenic refrigerator mainly consists of two apparatus, a helium compressor and a cold box which needs supply of cooling water. Therefore, cooling water system is essential to operate of cryogenic refrigerator normally. This report is mainly focused on the detailed design of the cooling water system for the HANARO cold neutron source, and describes design requirement, calculation, specification of equipment and water treatment method

Development of the interactive model between Component Cooling Water System and Containment Cooling System using GOTHIC

International Nuclear Information System (INIS)

Byun, Choong Sup; Song, Dong Soo; Jun, Hwang Yong

2006-01-01

In a design point of view, component cooling water (CCW) system is not full-interactively designed with its heat loads. Heat loads are calculated from the CCW design flow and temperature condition which is determined with conservatism. Then the CCW heat exchanger is sized by using total maximized heat loads from above calculation. This approach does not give the optimized performance results and the exact trends of CCW system and the loads during transient. Therefore a combined model for performance analysis of containment and the component cooling water(CCW) system is developed by using GOTHIC software code. The model is verified by using the design parameters of component cooling water heat exchanger and the heat loads during the recirculation mode of loss of coolant accident scenario. This model may be used for calculating the realistic containment response and CCW performance, and increasing the ultimate heat sink temperature limits

Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

1978-09-01

The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure

Water-cooled, fire boom blanket, test and evaluation for system prototype development

International Nuclear Information System (INIS)

Stahovec, J. G.; Urban, R. W.

1999-01-01

Initial development of actively cooled fire booms indicated that water-cooled barriers could withstand direct oil fire for several hours with little damage if cooling water were continuously supplied. Despite these early promising developments, it was realized that to build reliable full-scale system for Navy host salvage booms would require several development tests and lengthy evaluations. In this experiment several types of water-cooled fire blankets were tested at the Oil and Hazardous Materials Simulated Test Tank (OHMSETT). After the burn test the blankets were inspected for damage and additional tests were conducted to determine handling characteristics for deployment, recovery, cleaning and maintenance. Test results showed that water-cooled fire boom blankets can be used on conventional offshore oil containment booms to extend their use for controlling large floating-oil marine fires. Results also demonstrated the importance of using thermoset rubber coated fabrics in the host boom to maintain sufficient reserve seam strength at elevated temperatures. The suitability of passively cooled covers should be investigated to protect equipment and boom from indirect fire exposure. 1 ref., 2 tabs., 8 figs

The influence and analysis of natural crosswind on cooling characteristics of the high level water collecting natural draft wet cooling tower

Science.gov (United States)

Ma, Libin; Ren, Jianxing

2018-01-01

Large capacity and super large capacity thermal power is becoming the main force of energy and power industry in our country. The performance of cooling tower is related to the water temperature of circulating water, which has an important influence on the efficiency of power plant. The natural draft counter flow wet cooling tower is the most widely used cooling tower type at present, and the high cooling tower is a new cooling tower based on the natural ventilation counter flow wet cooling tower. In this paper, for high cooling tower, the application background of high cooling tower is briefly explained, and then the structure principle of conventional cooling tower and high cooling tower are introduced, and the difference between them is simply compared. Then, the influence of crosswind on cooling performance of high cooling tower under different wind speeds is introduced in detail. Through analysis and research, wind speed, wind cooling had little impact on the performance of high cooling tower; wind velocity, wind will destroy the tower inside and outside air flow, reducing the cooling performance of high cooling tower; Wind speed, high cooling performance of cooling tower has increased, but still lower than the wind speed.

Study of Cooling Characteristic of The Containment APWR Model Using Laminar Subcooled Water Film

International Nuclear Information System (INIS)

Diah Hidayanti; Aryadi Suwono; Nathanael P Tandian; Ari Darmawan Pasek; Efrizon Umar

2009-01-01

One of mechanism utilized by the next-generation pressurized water reactor for cooling its containment passively is gravitationally falling water spray cooling. This paper focuses on the characteristic study using Fluent 5/6 program for the case of the containment outer wall cooling by laminar sub-cooled water film. The cooling system characteristics which will be discussed consist of water film thickness and temperature on all parts of the containment wall as well as the effect of water spray volume flow rate on the water film thickness and convection heat transfer capability from the containment wall to the film bulk. In addition, some kinds of non dimensional numbers involved in the film heat transfer correlation will be presented in this paper. (author)

40 CFR 463.10 - Applicability; description of the contact cooling and heating water subcategory.

Science.gov (United States)

2010-07-01

... contact cooling and heating water subcategory. 463.10 Section 463.10 Protection of Environment... SOURCE CATEGORY Contact Cooling and Heating Water Subcategory § 463.10 Applicability; description of the contact cooling and heating water subcategory. This subpart applies to discharges of pollutants from...

Deep lake water cooling a renewable technology

Energy Technology Data Exchange (ETDEWEB)

Eliadis, C.

2003-06-01

In the face of increasing electrical demand for air conditioning, the damage to the ozone layer by CFCs used in conventional chillers, and efforts to reduce the greenhouse gases emitted into the atmosphere by coal-fired power generating stations more and more attention is focused on developing alternative strategies for sustainable energy. This article describes one such strategy, namely deep lake water cooling, of which the Enwave project recently completed on the north shore of Lake Ontario is a prime example. The Enwave Deep Lake Water Cooling (DLWC) project is a joint undertaking by Enwave and the City of Toronto. The $180 million project is unique in design and concept, using the coldness of the lake water from the depths of Lake Ontario (not the water itself) to provide environmentally friendly air conditioning to office towers. Concurrently, the system also provides improved quality raw cold water to the city's potable water supply. The plant has a rated capacity of 52,200 tons of refrigeration. The DLWC project is estimated to save 75-90 per cent of the electricity that would have been generated by a coal-fired power station. Enwave, established over 20 years ago, is North America's largest district energy system, delivering steam, hot water and chilled water to buildings from a central plant via an underground piping distribution network. 2 figs.

Device for preventing cooling water from flowing out of reactor

International Nuclear Information System (INIS)

Chinen, Masanori; Kotani, Koichi; Murase, Michio.

1976-01-01

Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

Development of a fast-water field guide

International Nuclear Information System (INIS)

Hansen, K.A.

2001-01-01

There are several manuals for oil spill response, but few have information on fast-water conditions. Between 1992 and 1997, approximately 58 per cent of all the oil spilled by volume in the United States happened in waterways with currents exceeding one knot, and the Coast Guard recognized the absence of standard terminology that could be used for fast-water responses. For that reason, an initiative was undertaken to create a document that addresses only fast-water issues. The resulting field guide can be used for training or responding to spills in fast-water. The user must rely on other manuals for issues on toxicity and shoreline cleanup as well as local contingency and site safety plans. The fast-water guide allows on-scene commanders and area supervisors the ability to define techniques and terminology for the responders in the field. It is particularly useful for Coast Guard Marine Safety Units when working with Coast Guard operational units during an emergency response. The current version of the guide that is under review by the working group contains 9 chapters and 9 appendices. The guide includes a decision-matrix that identifies various fat-water scenarios and provides recommended strategies. It then links to other sections of the document that contain details about the necessary equipment configurations. Photographs are provided to reinforce the concepts. The guide includes a checklist of the issues that must be addressed in any spill, such as weather and nature of the spill with some fast water issues added. Links to appropriate Internet sites are also included in the guide. Information within the guide can be condensed to one sheet for use in the field. 9 refs., 4 tabs., 11 figs

POOL WATER TREATMENT AND COOLING SYSTEM DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

King, V.

2000-01-01

The Pool Water Treatment and Cooling System is located in the Waste Handling Building (WHB), and is comprised of various process subsystems designed to support waste handling operations. This system maintains the pool water temperature within an acceptable range, maintains water quality standards that support remote underwater operations and prevent corrosion, detects leakage from the pool liner, provides the capability to remove debris from the pool, controls the pool water level, and helps limit radiological exposure to personnel. The pool structure and liner, pool lighting, and the fuel staging racks in the pool are not within the scope of the Pool Water Treatment and Cooling System. Pool water temperature control is accomplished by circulating the pool water through heat exchangers. Adequate circulation and mixing of the pool water is provided to prevent localized thermal hotspots in the pool. Treatment of the pool water is accomplished by a water treatment system that circulates the pool water through filters, and ion exchange units. These water treatment units remove radioactive and non-radioactive particulate and dissolved solids from the water, thereby providing the water clarity needed to conduct waste handling operations. The system also controls pool water chemistry to prevent advanced corrosion of the pool liner, pool components, and fuel assemblies. Removal of radioactivity from the pool water contributes to the project ALARA (as low as is reasonably achievable) goals. A leak detection system is provided to detect and alarm leaks through the pool liner. The pool level control system monitors the water level to ensure that the minimum water level required for adequate radiological shielding is maintained. Through interface with a demineralized water system, adequate makeup is provided to compensate for loss of water inventory through evaporation and waste handling operations. Interface with the Site Radiological Monitoring System provides continuous

Biofouling on Coated Carbon Steel in Cooling Water Cycles Using Brackish Seawater

Directory of Open Access Journals (Sweden)

Pauliina Rajala

2016-11-01

Full Text Available Water cooling utilizing natural waters is typically used for cooling large industrial facilities such as power plants. The cooling water cycles are susceptible to biofouling and scaling, which may reduce heat transfer capacity and enhance corrosion. The performance of two fouling-release coatings combined with hypochlorite treatment were studied in a power plant utilizing brackish sea water from the Baltic Sea for cooling. The effect of hypochlorite as an antifouling biocide on material performance and species composition of microfouling formed on coated surfaces was studied during the summer and autumn. Microfouling on surfaces of the studied fouling-release coatings was intensive in the cooling water cycle during the warm summer months. As in most cases in a natural water environment the fouling consisted of both inorganic fouling and biofouling. Chlorination decreased the bacterial number on the surfaces by 10–1000 fold, but the efficacy depended on the coating. In addition to decreasing the bacterial number, the chlorination also changed the microbial species composition, forming the biofilm on the surfaces of two fouling-release coatings. TeknoTar coating was proven to be more efficient in combination with the hypochlorite treatment against microfouling under these experimental conditions.

Ecological impact of chloro-organics produced by chlorination of cooling tower waters

International Nuclear Information System (INIS)

Jolley, R.L.; Cumming, R.B.; Pitt, W.W.; Taylor, F.G.; Thompson, J.E.; Hartmann, S.J.

1977-01-01

Experimental results of the initial assessment of chlorine-containing compounds in the blowdown from cooling towers and the possible mutagenic activity of these compounds are reported. High-resolution liquid chromatographic separations were made on concentrates of the blowdown from the cooling tower at the High Flux Isotope Reactor (HFIR) and from the recirculating water system for the cooling towers at the Oak Ridge Gaseous Diffusion Plant (ORGDP), Oak Ridge, Tennessee. The chromatograms of chlorinated cooling waters contained numerous uv-absorbing and cerate-oxidizable constituents that are now being processed through a multicomponent identification procedure. Concentrates of the chlorinated waters are also being examined for mutagenic activity

Use of Produced Water in Recirculated Cooling Systems at Power Generating Facilities

Energy Technology Data Exchange (ETDEWEB)

C. McGowin; M. DiFilippo; L. Weintraub

2006-06-30

Tree ring studies indicate that, for the greater part of the last three decades, New Mexico has been relatively 'wet' compared to the long-term historical norm. However, during the last several years, New Mexico has experienced a severe drought. Some researchers are predicting a return of very dry weather over the next 30 to 40 years. Concern over the drought has spurred interest in evaluating the use of otherwise unusable saline waters to supplement current fresh water supplies for power plant operation and cooling and other uses. The U.S. Department of Energy's National Energy Technology Laboratory sponsored three related assessments of water supplies in the San Juan Basin area of the four-corner intersection of Utah, Colorado, Arizona, and New Mexico. These were (1) an assessment of using water produced with oil and gas as a supplemental supply for the San Juan Generating Station (SJGS); (2) a field evaluation of the wet-surface air cooling (WSAC) system at SJGS; and (3) the development of a ZeroNet systems analysis module and an application of the Watershed Risk Management Framework (WARMF) to evaluate a range of water shortage management plans. The study of the possible use of produced water at SJGS showed that produce water must be treated to justify its use in any reasonable quantity at SJGS. The study identified produced water volume and quality, the infrastructure needed to deliver it to SJGS, treatment requirements, and delivery and treatment economics. A number of produced water treatment alternatives that use off-the-shelf technology were evaluated along with the equipment needed for water treatment at SJGS. Wet surface air-cooling (WSAC) technology was tested at the San Juan Generating Station (SJGS) to determine its capacity to cool power plant circulating water using degraded water. WSAC is a commercial cooling technology and has been used for many years to cool and/or condense process fluids. The purpose of the pilot test was to

Fast cooling phenomena with ice pellet injection in the JIPP T-IIU Tokamak

International Nuclear Information System (INIS)

Sakamoto, M.; Sato, K.N.; Ogawa, Y.

1991-01-01

Ice pellet injection experiments were carried out in the JIPP T-IIU Tokamak in order to study thermal (cooling) transport just after injection. The cut-off problem of ECE signals due to the rise in density has been resolved by careful measurements of the temperature profile at a high time resolution (Δt = 2μs) during its decay phase. The phenomenon of ultra-fast cooling (so-called pre-cooling) has been identified using the two different methods of ECE and soft X-ray (SXR) measurements. In the outer region (r > r q=1 ) of the plasma the cooling propagation velocity is comparable to or slightly greater than the pellet velocity, while in the central region (r q=1 ) the propagation velocity is significantly greater than the pellet velocity. Ice pellets were injected into various kinds of JIPP T-IIU plasmas, the current and sawtooth phase of which had different values, including a no-sawtooth plasma. The existence of the q = 1 surface and arrival of a pellet near the q = 1 surface have turned out to be necessary conditions for pre-cooling, and even just after the sawtooth crash the pre-cooling starts around the q = 1 surface, not at the plasma center. Simultaneous measurements of electron temperature and density profiles indicate that the central temperature always decreases before the central density increases. Some anomalous transport might be induced by pellet injection at the central region. (Author)

The cooling water from Ringhals

International Nuclear Information System (INIS)

1980-10-01

The Ringhals Nuclear Power Plant is situated on the Swedish west coast about 70 km south of Gothenburg. At present two units operate at a total maximum power level of 1580 MWE and their once-through cooling system requires 80 m 3 /sec sea water. The temperature of the cooling water increases approximately 10 deg C. This study assesses the spreading of the discharged cooling water in the ambient sea and is based on field data sampled since the end of 1974. About 50 thermal mappings were made in the area by boat or in some cases by aeroplane. Several continously recording current and temperature instruments were used. Water samples analysed for salinity, oxygen and turbidity were collected most of the time. Through the thermal mappings four main directions of the thermal plume were distinguished: northward along the coast (class 1A), northward further out (class 1B), westward and reversing plumes (class 2) and southward (class 3). The changing of the plume hour by hour between these main directions was measured by the recording temperature instruments. Data from almost one year gave the following statistics: 40 percent class 1A + 1B, 15 percent class 2, 25 percent class 3 and 20 percent undefined directions. Furthermore, available data showed that the direction of the ambient current mostly gave the plume direction. The wind, on the other hand, was more uncertain as an indicator of the plume direction. Owing to the varying ambient currents the plume changed its direction more than once a day. Measurable excess temperatures were found within a few kilometers wide zone from Stavder in the north to Norra Horta in the south. The largest measured area with excess temperatures of more than 1 deg C was 6 km 2 . Usually, however, the plume covered about 2.5 km 2 at full production at the power plant. As for the downward spreading, the bottom of the plume normally registrated down to 3-7 m, but occasionally it reached the 10 - 12 m level. The tendency of deep penetration

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

ETDR, The European Union's Experimental Gas-Cooled Fast Reactor Project

International Nuclear Information System (INIS)

Poette, Christian; Brun-Magaud, Valerie; Morin, Franck; Dor, Isabelle; Pignatel, Jean-Francois; Bertrand, Frederic; Stainsby, Richard; Pelloni, Sandro; Every, Denis; Da Cruz, Dirceu

2008-01-01

In the Gas-Cooled Fast Reactor (GFR) development plan, the Experimental Technology Demonstration Reactor (ETDR) is the first necessary step towards the electricity generating prototype GFR. It is a low power (∼50 MWth) Helium cooled fast reactor. The pre-conceptual design of the ETDR is shared between European partners through the GCFR Specifically Targeted Research Project (STREP) within the European Commission's 6. R and D Framework Program. After recalling the place of ETDR in the GFR development plan, the main reactor objectives, the role of the European partners in the different design and safety tasks, the paper will give an overview of the current design with recent progresses in various areas like: - Sub-assembly technology for the starting core (pin bundle with MOX fuel and stainless steel cladding). - The design of experimental advanced ceramic GFR fuel sub-assemblies included in several locations of the starting core. - Starting Core reactivity management studies model including experimental GFR sub-assemblies. - Neutron and radiation shielding calculations using a specific MCNP model. The model allows evaluation of the neutron doses for the vessel and internals and radiation doses for maintenance operations. - System design and safety considerations, with a reactor architecture largely influenced by the Decay Heat Removal strategy (DHR) for de-pressurized accidents. The design of the reactor raises a number of issues in terms of fuel, neutronics, thermal-hydraulics codes qualification as well as critical components (blowers, IHX, thermal barriers) qualification. An overview of the R and D development on codes and technology qualification program is presented. Finally, the status of international collaborations and their perspectives for the ETDR are mentioned. (authors)

Improving of the photovoltaic / thermal system performance using water cooling technique

International Nuclear Information System (INIS)

Hussien, Hashim A; Numan, Ali H; Abdulmunem, Abdulmunem R

2015-01-01

This work is devoted to improving the electrical efficiency by reducing the rate of thermal energy of a photovoltaic/thermal system (PV/T).This is achieved by design cooling technique which consists of a heat exchanger and water circulating pipes placed at PV module rear surface to solve the problem of the high heat stored inside the PV cells during the operation. An experimental rig is designed to investigate and evaluate PV module performance with the proposed cooling technique. This cooling technique is the first work in Iraq to dissipate the heat from PV module. The experimental results indicated that due to the heat loss by convection between water and the PV panel's upper surface, an increase of output power is achieved. It was found that without active cooling, the temperature of the PV module was high and solar cells could only achieve a conversion efficiency of about 8%. However, when the PV module was operated under active water cooling condition, the temperature was dropped from 76.8°C without cooling to 70.1°C with active cooling. This temperature dropping led to increase in the electrical efficiency of solar panel to 9.8% at optimum mass flow rate (0.2L/s) and thermal efficiency to (12.3%). (paper)

State waste discharge permit application for cooling water and condensate discharges

Energy Technology Data Exchange (ETDEWEB)

Haggard, R.D.

1996-08-12

The following presents the Categorical State Waste Discharge Permit (SWDP) Application for the Cooling Water and Condensate Discharges on the Hanford Site. This application is intended to cover existing cooling water and condensate discharges as well as similar future discharges meeting the criteria set forth in this document.

Absorption cooling sources atmospheric emissions decrease by implementation of simple algorithm for limiting temperature of cooling water

Science.gov (United States)

Wojdyga, Krzysztof; Malicki, Marcin

2017-11-01

Constant strive to improve the energy efficiency forces carrying out activities aimed at reduction of energy consumption hence decreasing amount of contamination emissions to atmosphere. Cooling demand, both for air-conditioning and process cooling, plays an increasingly important role in the balance of Polish electricity generation and distribution system in summer. During recent years' demand for electricity during summer months has been steadily and significantly increasing leading to deficits of energy availability during particularly hot periods. This causes growing importance and interest in trigeneration power generation sources and heat recovery systems producing chilled water. Key component of such system is thermally driven chiller, mostly absorption, based on lithium-bromide and water mixture. Absorption cooling systems also exist in Poland as stand-alone systems, supplied with heating from various sources, generated solely for them or recovered as waste or useless energy. The publication presents a simple algorithm, designed to reduce the amount of heat for the supply of absorption chillers producing chilled water for the purposes of air conditioning by reducing the temperature of the cooling water, and its impact on decreasing emissions of harmful substances into the atmosphere. Scale of environmental advantages has been rated for specific sources what enabled evaluation and estimation of simple algorithm implementation to sources existing nationally.

Methods for the sodium cooled fast reactor fire safety provisions

International Nuclear Information System (INIS)

Gryaznov, B.V.; Dergachev, N.P.

1983-01-01

Problems of fire safety provision on NPPs with sodium cooled fast reactor are under discussion. Methods of sodium leak localization, measures eliminating sodium flaring up during leaks and main means of sodium fire extinguishing are considered. An extinguishing of sodium flaring up is performed by means of sodium temperatUre decrease and by limitation of hydrogen access to the flaring up surface. A conclusion is made that the most effective methods of extinguishing are the following: self-extinguishing (due to hydrogen burning out in a limiting volume); extinguishing by a gas mixture of nitrogen and carbonic acid (initial filling and blowing of rooms during sodium flaring up); extinguishing by special powders

Study on flow-induced vibration of large-diameter pipings in a sodium-cooled fast reactor. Influence of elbow curvature on velocity fluctuation field

International Nuclear Information System (INIS)

Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira

2010-02-01

The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9m/s in the primary hot leg pipe with large-diameter (1.3m). The maximum Reynolds number in the piping reaches 4.2x10 7 . The hot leg pipe having a 90 degree elbow with curvature ratio of r/D=1.0, so-called 'short elbow', which enables a compact reactor vessel. In sodium cooled fast reactors, the system pressure is so low that thickness of pipings in the cooling system is thinner than that in LWRs. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned at the short elbow. The evaluation of the structural integrity of pipings in JSFR should be conducted based on a mechanistic approach of FIV at the elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow. (author)

Study of various Brayton cycle designs for small modular sodium-cooled fast reactor

International Nuclear Information System (INIS)

Ahn, Yoonhan; Lee, Jeong Ik

2014-01-01

Highlights: • Application of closed Brayton cycle for small and medium sized SFRs is reviewed. • S-CO 2 , helium and nitrogen cycle designs for small modular SFR applications are analyzed and compared in terms of cycle efficiency, component performance and physical size. • Several new layouts for each Brayton cycle are suggested to simplify the turbomachinery designs. • S-CO 2 cycle design shows the best efficiency and compact size compared to other Brayton cycles. - Abstract: Many previous sodium cooled fast reactors (SFRs) adopted steam Rankine cycle as the power conversion system. However, the concern of sodium water reaction has been one of the major design issues of a SFR system. As an alternative to the steam Rankine cycle, several closed Brayton cycles including supercritical CO 2 cycle, helium cycle and nitrogen cycle have been suggested recently. In this paper, these alternative gas Brayton cycles will be compared to each other in terms of cycle performance and physical size for small modular SFR application. Several new layouts are suggested for each fluid while considering the turbomachinery design and the total system volume

Lead-cooled fast-neutron reactor (BREST) (Approaches to the closed NFC) - 5435

International Nuclear Information System (INIS)

Dragunov, Y.G.; Lemekhov, V.V.; Moiseyev, A.V.; Smirnov, V.S.; Tocheny, L.V.; Umanskiy, A.A.

2015-01-01

The BREST-OD-300 reactor is under development in Russia. It is an intrinsically safe pilot demonstration lead-cooled fast reactor with uranium-plutonium nitride fuel. This reactor is based on a new concept of inherent safety whose basic principles are: -) the exclusion of severe accidents at the plant (reactivity type, loss of cooling, fires, explosions) that require the resettlement of the population; -) the closing of the nuclear fuel cycle through the burning of minor actinides; -) the environmental acceptability through the maximal reduction of the amount of high-level long-lived radioactive waste nuclides - nuclear fuel cycle products, sent for the final disposal; -) the technological strengthening of non-proliferation. Closed fuel cycle with reactors of BREST type burning minor actinides gives the opportunity to achieve the radiation equivalence between radioactive wastes and natural uranium during a time period about 300 years

Organohalogens in chlorinated cooling waters discharged from nuclear power stations

International Nuclear Information System (INIS)

Bean, R.M.; Mann, D.C.; Neitzel, D.A.

1983-01-01

For the power plant discharges studied to date, measured concentrations of trihalomethanes are lower than might be expected, particularly in cooling tower water, which can lose THMs to the atmosphere. In the cooling towers, where chlorine was added in higher concentrations and for longer residence times, halogenated phenols can contribute significantly to the total organic halogen content of the discharge. The way in which cooling towers are operated may also influence the production of halogenated phenols because they concentrate the incoming water by a factor of 4 or 5. In addition, the phenols, which act as a substrate for the halogenating agent, are also probably concentrated by the cooling tower operation and may be prevented from being biodegraded by addition of the same biocide that produces the halogenated phenols. 8 references, 4 tables

Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

International Nuclear Information System (INIS)

Yoon, K. H.; Lee, C. B.

2014-01-01

The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

Experimental Progress in Fast Cooling in the ESR

CERN Document Server

Steck, Markus; Beller, Peter; Franzke, Bernhard; Nolden, Fritz

2005-01-01

The ESR storage ring at GSI is operated with highly charged heavy ions. Due to the high electric charge the ions interact much stronger with electromagnetic fields. Therefore both cooling methods which are applied to stored ions in the ESR, stochastic cooling and electron cooling, are more powerful than for singly charged particles. The experimental results exhibit cooling times for stochastic cooling of a few seconds. For cold ion beams, electron cooling provides cooling times which are one to two orders of magnitude smaller. The beams are cooled to beam parameters which are limited by intrabeam scattering. At small ion numbers, however, intrabeam scattering is suppressed by electron cooling, clear evidence was found that the ion beam forms a one-dimensional ordered structure, a linear chain of ions. The strengths of stochastic cooling and electron cooling are complementary and can be combined favorably. Stochastic cooling is employed for pre-cooling of hot secondary beams followed by electron cooling to pro...

The role of the IAEA in advanced technologies for water-cooled reactors

International Nuclear Information System (INIS)

Cleveland, J.

1996-01-01

The role of the IAEA in advanced technologies for water-cooled reactors is described, including the following issues: international collaboration ways through international working group activities; IAEA coordinated research programmes; cooperative research in advanced water-cooled reactor technology

Biological effects from discharge of cooling water from thermal power plants

International Nuclear Information System (INIS)

1976-12-01

Results are reported for a Danish project on biological effects from discharge of cooling water from thermal power plants. The purpose of the project was to provide an up-to-date knowledge of biological effects of cooling water discharge and of organization and evaluation of recipient investigations in planned and established areas. (BP)

Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

1982-01-01

Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

Corrosion evaluation of cooling-water treatments for gas centrifuge facilities

International Nuclear Information System (INIS)

Schmidt, C.R.; Meredith, P.F.

1980-01-01

The corrosion resistance of six different types of weighted metal coupons was evaluated at 29 0 C (84 0 F) in flowing water containing nitrite-borate-silicate corrosion inhibitors. The question for evaluation was whether it would be more advantageous: (1) to drain the treated cooling water from the centrifuge machine and to expose them to moisture-laden air over an assumed shop downtime and repair perid of 1 month; or (2) to let the treated cooling water remain stagnant in the machines during this downtime. The moisture-laden-air exposure was more detrimental

Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

Directory of Open Access Journals (Sweden)

Jaewoon Yoo

2016-10-01

Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

International Nuclear Information System (INIS)

Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

2016-01-01

The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

OpenAIRE

Verma, Vasudha

2017-01-01

Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

International Nuclear Information System (INIS)

2012-03-01

The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

Water cooling of RF structures

International Nuclear Information System (INIS)

Battersby, G.; Zach, M.

1994-06-01

We present computer codes for heat transfer in water cooled rf cavities. RF parameters obtained by SUPERFISH or analytically are operated on by a set of codes using PLOTDATA, a command-driven program developed and distributed by TRIUMF [1]. Emphasis is on practical solutions with designer's interactive input during the computations. Results presented in summary printouts and graphs include the temperature, flow, and pressure data. (authors). 4 refs., 4 figs

Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

International Nuclear Information System (INIS)

Tsoulfanidis, N.; Jankhah, M.H.

1979-01-01

The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

Directory of Open Access Journals (Sweden)

Guidez Joel

2017-01-01

In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

LOFA analyses for the water and helium cooled SEAFP reactors

International Nuclear Information System (INIS)

Sponton, L.; Sjoeberg, A.; Nordlinder, S.

2001-01-01

This study was performed in the frame of the European long-term fusion safety programme 1999 (SEAFP99). Loss of flow accidents (LOFA) have been studied for two cases, first for a helium cooled reactor with advanced dual-coolant (DUAL) blanket at 100% nominal power. The second case applies to a water-cooled reactor at 20% nominal power. Both transients were simulated with the code MELCOR 1.8.4. The results for the helium cooled reactor show that with a natural circulation flow of helium after the pump stops, the first wall temperature will stay below the temperature for excepted failure of the construction material. For the water cooled reactor, the results show that the pressurizer set point for its liquid volumetric inventory is reached before the plasma facing components attain a critical temperature. The pressurizer set point will induce a plasma shutdown

European supercritical water cooled reactor

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

2011-01-01

Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

Energy Technology Data Exchange (ETDEWEB)

Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

2016-05-15

Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

Reprocessing technology of liquid metal cooled fast breeder reactor fuel

International Nuclear Information System (INIS)

Baetsle, L.H.; Broothaerts, J.; Heylen, P.R.; Eschrich, H.; Geel, J. van

1974-11-01

All the important aspects of LMFBR fuel reprocessing are critically reviewed in this report. Storage and transportation techniques using sodium, inert gas, lead, molten salts and organic coolants are comparatively discussed in connection with cooling time and de-activation techniques. Decladding and fuel disaggregation of UO 2 -PuO 2 fuel are reviewed according to the present state of R and D in the main nuclear powers. Strong emphasis is put on on voloxidation, mechanical pulverization and molten salt disaggregation in connection with volatilization of gaseous fission products. Release of fission gases and the resulting off-gas treatment are discussed in connection with cooling time, burn up and dissagregation techniques. The review is limited to tritium, iodine xenon-krypton and radioactive airborne particulates. Dissolution, solvent extraction and plutonium purification problems specifically connected to LMFBR fuel are reviewed with emphasis on the differences between LWR and fast fuel reprocessing. Finally the categories of wastes produced by reprocessing are analysed according to their origin in the plant and their alpha emitters content. The suitable waste treatment techniques are discussed in connection with the nature of the wastes and the ultimate disposal technique. (author)

Thermodynamic analysis of cooling systems for nuclear power stations condenser

International Nuclear Information System (INIS)

Beck, A.

1985-06-01

This work is an attempt to concentrate on the thermodynamic theory, the engineering solution and the quantities of water needed for the operation of a wet as well as a wet/dry cooling towers coupled to a nuclear turbine condenser,. About two hundred variables are needed for the design of a condenser - cooling tower system. In order to make the solution fast and handy, a computer model was developed. The amount of water evaporation from cooling towers is a function of the climate conditions prevailing around the site. To achieve an authentic analysis, the meteorological data of the northern Negev was used. The total amount of water necessary to add to the system in a year time of operation is large and is a function of both the blow-down rate and the evaporation. First estimations show that the use of a combined system, wet/dry cooling tower, is beneficial in the northern Negev area. Such a system can reduce significantly the amount of wasted fresh water. Lack of international experience is the major problem in the acceptability of wet/dry cooling towers. The technology of a wet cooling tower using sea water is also discussed where no technical or engineering limitations were found. This work is an attempt to give some handy tools for making the choice of cooling systems for nuclear power plants easier

Upgrade of the cooling water temperature measures system for HLS

International Nuclear Information System (INIS)

Guo Weiqun; Liu Gongfa; Bao Xun; Jiang Siyuan; Li Weimin; He Duohui

2007-01-01

The cooling water temperature measures system for HLS (Hefei Light Source) adopts EPICS to the developing platform and takes the intelligence temperature cruise instrument for the front control instrument. Data of temperatures are required by IOCs through Serial Port Communication, archived and searched by Channel Archiver. The system can monitor the real-time temperatures of many channels cooling water and has the function of history data storage, and data network search. (authors)

77 FR 73056 - Initial Test Programs for Water-Cooled Nuclear Power Plants

Science.gov (United States)

2012-12-07

... Plants AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide; request for comment... (DG), DG-1259, ``Initial Test Programs for Water-Cooled Nuclear Power Plants.'' This guide describes... (ITPs) for light water cooled nuclear power plants. DATES: Submit comments by January 31, 2013. Comments...

Economic competitiveness requirements for evolutionary water cooled reactors

International Nuclear Information System (INIS)

Hudson, C.R.; Bertel, E.; Paik, K.H.; Roh, J.H.; Tort, V.

1999-01-01

This paper analyses the necessary economic conditions for evolutionary water cooled reactors to be competitive. Utilising recent national cost data for fossil-fired base load plants expected to be commissioned by 2005 -2010, target costs for nuclear power plants are discussed. Factors that could contribute to the achievement of those targets by evolutionary water cooled reactors are addressed. The feed-back from experience acquired in implementing nuclear programmes is illustrated by some examples from France and the Republic of Korea. The paper discusses the impacts on nuclear power competitiveness of globalisation and deregulation of the electricity market and privatisation of the electricity sector. In addition, issues related to external cost internalisation are considered. (author)

Sustained Recycle in Light Water and Sodium-Cooled Reactors

International Nuclear Information System (INIS)

Piet, Steven J.; Bays, Samuel E.; Pope, Michael A.; Youinou, Gilles J.

2010-01-01

From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

Impact of nuclear data on sodium-cooled fast reactor calculations

International Nuclear Information System (INIS)

Aures, A.; Bostelmann, F.; Zwermann, W.; Velkov, K.

2016-01-01

Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors. (authors)

Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

International Nuclear Information System (INIS)

Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

2014-01-01

Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

Design measures in evolutionary water cooled reactors to optimize for economic viability

International Nuclear Information System (INIS)

Oh, S.J.; Yu, S.K.W.; Appell, B.

1999-01-01

Since the mid 1980s, there have been various efforts to develop evolutionary water cooled reactors based on the current operating plant experience. To sustain and improve the economic viability, particular attention has been paid to the following aspects in developing evolutionary water cooled reactors: design simplification and increased operating margins, standardization in design as well as construction and operation, integration of operating plant insights, and consideration of safety, operability and constructability during the design stage. This paper reviews each item and discusses several examples from some of the evolutionary water cooled reactors being developed. (author)

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

CFD results for temperature dependence water cooling pump NPSH calculations - 15425

International Nuclear Information System (INIS)

Strongin, M.P.

2015-01-01

In this work the possibility to model the pump for water cooling reactors behavior in the critical situation was considered for cases when water temperature suddenly increases. In cases like this, cavitation effects may cause pump shutoff and consequently stop the reactor cooling. Centrifugal pump was modeled. The calculations demonstrate strong dependence of NPSH (net-positive-suction-head) on the water temperature on the pump inlet. The water temperature on the inlet lies between 25 and 180 C. degrees. The pump head performance curve has a step-like slope below NPSH point. Therefore, if the pressure on the pump inlet is below than NPSH, it leads to the pump shutoff. For high water temperature on the pump inlet, NPSH follows the vapor saturated pressure for given temperature with some offset. The results clearly show that in case of accidental increase of temperature in the cooling loop, special measures are needed to support the pressure on the pump inlet to prevent pump shutoff. (author)

A three-dimensional thermal and fluid dynamics analysis of a gas cooled subcritical fast reactor driven by a D-T fusion neutron source

International Nuclear Information System (INIS)

Angelo, G.; Andrade, D.A.; Angelo, E.; Carluccio, T.; Rossi, P.C.R.; Talamo, A.

2011-01-01

Highlights: → A thermal fluid dynamics numerical model was created for a gas cooled subcritical fast reactor. → Standard k-ε model, Eddy Viscosity Transport Equation model underestimates the fuel temperature. → For a conservative assumption, SSG Reynolds stress model was chosen. → Creep strength is the most important parameter in fuel design. - Abstract: The entire nuclear fuel cycle involves partitioning classification and transmutation recycling. The usage of a tokamak as neutron sources to burn spent fuel in a gas cooled subcritical fast reactor (GCSFR) reduces the amount of long-lived radionuclide, thus increasing the repository capacity. This paper presents numerical thermal and fluid dynamics analysis for a gas cooled subcritical fast reactor. The analysis aim to determine the operational flow condition for this reactor, and to compare three distinct turbulence models (Eddy Viscosity Transport Equation, standard k-ε and SSG Reynolds stress) for this application. The model results are presented and discussed. The methodology used in this paper was developed to predict the coolant mass flow rate. It can be applied to any other gas cooled reactor.

Gas-Cooled Fast Reactor (GFR) Decay Heat Removal Concepts

International Nuclear Information System (INIS)

K. D. Weaver; L-Y. Cheng; H. Ludewig; J. Jo

2005-01-01

Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report is a compilation of work performed on decay heat removal systems for a 2400 MWt GFR during this fiscal year (FY05)

Engineering and economic evaluation of wet/dry cooling towers for water conservation

International Nuclear Information System (INIS)

Hu, M.C.

1976-11-01

The results are presented of a design and cost study for wet/dry tower systems used in conjunction with 1000 MWe nuclear power plants to reject waste heat while conserving water. Design and cost information for wet/dry tower systems are presented, and these cooling system alternatives are compared with wet and dry tower systems to determine whether the wet/dry tower concept is an economically viable alternative. The wet/dry cooling tower concept investigated is one which combines physically separated wet towers and dry towers into an operational unit. In designing the wet/dry tower, a dry cooling tower is sized to carry the plant heat load at low ambient temperatures, and a separate wet tower is added to augment the heat rejection of the dry tower at higher ambient temperatures. These wet/dry towers are designed to operate with a conventional low back pressure turbine commercially available today. The component wet and dry towers are state-of-the-art designs. From this study it was concluded that: wet/dry cooling systems can be designed to provide a significant economic advantage over dry cooling yet closely matching the dry tower's ability to conserve water, a wet/dry system which saves as much as 99 percent of the make-up water required by a wet tower can maintain that economic advantage, and therefore, for power plant sites where water is in short supply, wet/dry cooling is the economic choice over dry cooling

COGNITIVE AND PHYSIOLOGICAL INITIAL RESPONSES DURING COOL WATER IMMERSION

Directory of Open Access Journals (Sweden)

Alex Buoite Stella

2014-12-01

Full Text Available The initial responses during water immersion are the first mechanisms reacting to a strong stimulation of superficial nervous cold receptors. Cold shock induces tachycardia, hypertension, tachypnea, hyperventilation, and reduced end-tidal carbon dioxide fraction. These initial responses are observed immediately after the immersion, they last for about 3 min and have been also reported in water temperatures up to 25 °C. the aim of the present study was to observe cognitive and physiological functions during immersion in water at cool temperature. Oxygen consumption, ventilation, respiratory frequency, heart rate and expired fraction of oxygen were measured during the experiment. A code substitution test was used to evaluate executive functions and, specifically, working memory. This cognitive test was repeated consecutively 6 times, for a total duration of 5 minutes. Healthy volunteers (n = 9 performed the test twice in a random order, once in a dry thermoneutral environment and once while immersed head-out in 18 °C water. The results indicated that all the physiological parameters were increased during cool water immersion when compared with the dry thermoneutral condition (p < 0.05. Cognitive performance was reduced during the cool water immersion when compared to the control condition only during the first 2 min (p < 0.05. Our results suggest that planning the best rescue strategy could be partially impaired not only because of panic, but also because of the cold shock.

Environmental effects of large discharges of cooling water. Experiences from Swedish nuclear power plants

International Nuclear Information System (INIS)

Ehlin, Ulf; Lindahl, Sture; Neuman, Erik; Sandstroem, Olof; Svensson, Jonny

2009-07-01

Monitoring the environmental effects of cooling water intake and discharge from Swedish nuclear power stations started at the beginning of the 1960s and continues to this day. In parallel with long-term monitoring, research has provided new knowledge and methods to optimise possible discharge locations and design, and given the ability to forecast their environmental effects. Investigations into the environmental effects of cooling-water are a prerequisite for the issuing of power station operating permits by the environmental authorities. Research projects have been carried out by scientists at universities, while the Swedish Environmental Protection Agency, the Swedish Board of Fisheries, and the Swedish Meteorological and Hydrological Institute, SMHI, are responsible for the greater part of the investigations as well as of the research work. The four nuclear power plants dealt with in this report are Oskarshamn, Ringhals, Barsebaeck and Forsmark. They were taken into operation in 1972, 1975, 1975 and 1980 resp. - a total of 12 reactors. After the closure of the Barsebaeck plants in 2005, ten reactors remain in service. The maximum cooling water discharge from the respective stations was 115, 165, 50 and 135 m 3 /s, which is comparable to the mean flow of an average Swedish river - c:a 150 m 3 /s. The report summarizes studies into the consequences of cooling water intake and discharge. Radiological investigations made at the plants are not covered by this review. The strategy for the investigations was elaborated already at the beginning of the 1960s. The investigations were divided into pre-studies, baseline investigations and monitoring of effects. Pre-studies were partly to gather information for the technical planning and design of cooling water intake and outlet constructions, and partly to survey the hydrographic and ecological situation in the area. Baseline investigations were to carefully map the hydrography and ecology in the area and their natural

System for fast and accurate filling of a two-phase cooling device, notably a heat pipe, adapted for use in an automated process

NARCIS (Netherlands)

Wits, Wessel Willems; Ten Hoeve, Harm Jan; Te Riele Gert, Jan; Van Es, Johannes

2013-01-01

The current invention relates to a system for fast and accurate filling of a two-phase cooling device, comprising a binding device (30) intended to be hermetically mounted onto the cooling device, the binding device (30) comprising a through-hole (32) able to be in fluid contact with the cooling

SYSTEM FOR FAST AND ACCURATE FILLING OF A TWO-PHASE COOLING DEVICE, NOTABLY A HEAT PIPE, ADAPTED FOR USE IN AN AUTOMATED PROCESS

NARCIS (Netherlands)

Wits, Wessel Willems; Ten Hoeve, Harm Jan; Te Riele Gert, Jan; Van Es, Johannes

2013-01-01

The current invention relates to a system for fast and accurate filling of a two- phase cooling device, comprising a binding device (30) intended to be hermetically mounted onto the cooling device, the binding device (30) comprising a through-hole (32) able to be in fluid contact with the cooling

SYSTEM FOR FAST AND ACCURATE FILLING OF A TWO-PHASE COOLING DEVICE, NOTABLY A HEAT PIPE, ADAPTED FOR USE IN AN AUTOMATED PROCESS

NARCIS (Netherlands)

Wits, Wessel Willems; Ten Hoeve, Harm Jan; Te Riele Gert, Jan; Van Es, Johannes; Wits, Wessel Willems; Ten Hoeve, Harm Jan; Te Riele, Gerhardus Wilhelmus; Van Es, Johannes

2014-01-01

The current invention relates to a system for fast and accurate filling of a two-phase cooling device, comprising a binding device (30) intended to be hermetically mounted onto the cooling device, the binding device (30) comprising a through-hole (32) able to be in fluid contact with the cooling

Numerically Analysed Thermal Condition of Hearth Rollers with the Water-Cooled Shaft

Directory of Open Access Journals (Sweden)

A. V. Ivanov

2016-01-01

Full Text Available Continuous furnaces with roller hearth have wide application in the steel industry. Typically, furnaces with roller hearth belong to the class of medium-temperature heat treatment furnaces, but can be used to heat the billets for rolling. In this case, the furnaces belong to the class of high temperature heating furnaces, and their efficiency depends significantly on the reliability of the roller hearth furnace. In the high temperature heating furnaces are used three types of watercooled shaft rollers, namely rollers without insulation, rollers with insulating screens placed between the barrel and the shaft, and rollers with bulk insulation. The definition of the operating conditions of rollers with water-cooled shaft greatly facilitates the choice of their design parameters when designing. In this regard, at the design stage of the furnace with roller hearth, it is important to have information about the temperature distribution in the body of the rollers at various operating conditions. The article presents the research results of the temperature field of the hearth rollers of metallurgical heating furnaces. Modeling of stationary heat exchange between the oven atmosphere and a surface of rollers, and between the cooling water and shaft was executed by finite elements method. Temperature fields in the water-cooled shaft rollers of various designs are explored. The water-cooled shaft rollers without isolation, rollers with screen and rollers with bulk insulation, placed between the barrel and the water-cooled shaft were investigated. Determined the change of the thermo-physic parameters of the coolant, the temperature change of water when flowing in a pipe and shaft, as well as the desired pressure to supply water with a specified flow rate. Heat transfer coefficients between the cooling water and the shaft were determined directly during the solution based on the specified boundary conditions. Found that the greatest heat losses occur in the

Development in cooling water intake and outfall systems for atomic or steam power stations

International Nuclear Information System (INIS)

Wada, Akira

1987-01-01

The condenser cooling water channel, in its functional aspects, is an important structure for securing a stable supply of cooling water. In its design it is necessary to give a thorough-going study to a reduction of ranges affected by discharged warm water and minimizing the effect of discharged water on navigating ships, and in its functional aspects as a structure for power generation, avoiding the recirculation of discharged warm water as well as to maintaining the operation of power stations in case of abnormalities (concentration of dirts owing to typhoons and floods, outbreak of a large amount of jellyfishes, etc.), and all these aspects must be reflected in the design of cooling water channel systems. In this paper, the present situation relating to the design of cooling water intake and outfall systems in Japan is discussed. (author). 10 figs

Experiences with electrochemical analysis of copper at the PPB-level in saline cooling water and in the water/steam cycle

Energy Technology Data Exchange (ETDEWEB)

Thomsen, K [I/S Nordjyllandsvaerket, Vodskov (Denmark)

1996-12-01

Determination of trace amounts of copper in saline cooling water and in process water by differential pulse anodic stripping voltammetry combined with an UV-photolysis pretreatment is described. Copper concentrations well below 1 {mu}g/L may be analysed with a precision in the order of 10% and a high degree of accuracy. The basic principles of the method are described together with three applications covering analysis of cooling and process water samples. The analysis method has been applied to document the adherence of environmental limits for the copper uptake of cooling water passing brass condensers, to monitor the formation of protective layers of iron oxides on the cooling water side of brass condensers, and to study the transport of copper in water/steam cycles with heat exchangers and condensers of brass materials. (au)

New Mexico cloud super cooled liquid water survey final report 2009.

Energy Technology Data Exchange (ETDEWEB)

Beavis, Nick; Roskovensky, John K.; Ivey, Mark D.

2010-02-01

Los Alamos and Sandia National Laboratories are partners in an effort to survey the super-cooled liquid water in clouds over the state of New Mexico in a project sponsored by the New Mexico Small Business Assistance Program. This report summarizes the scientific work performed at Sandia National Laboratories during the 2009. In this second year of the project a practical methodology for estimating cloud super-cooled liquid water was created. This was accomplished through the analysis of certain MODIS sensor satellite derived cloud products and vetted parameterizations techniques. A software code was developed to analyze multiple cases automatically. The eighty-one storm events identified in the previous year effort from 2006-2007 were again the focus. Six derived MODIS products were obtained first through careful MODIS image evaluation. Both cloud and clear-sky properties from this dataset were determined over New Mexico. Sensitivity studies were performed that identified the parameters which most influenced the estimation of cloud super-cooled liquid water. Limited validation was undertaken to ensure the soundness of the cloud super-cooled estimates. Finally, a path forward was formulized to insure the successful completion of the initial scientific goals which include analyzing different of annual datasets, validation of the developed algorithm, and the creation of a user-friendly and interactive tool for estimating cloud super-cooled liquid water.

Counter flow induced draft cooling tower option for supercritical carbon dioxide Brayton cycle

Energy Technology Data Exchange (ETDEWEB)

Pidaparti, Sandeep R., E-mail: sandeep.pidaparti@gmail.com [Georgia Institute of Technology, George W. Woodruff School of Mechanical Engineering, Atlanta, GA 30332 (United States); Moisseytsev, Anton; Sienicki, James J. [Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Ranjan, Devesh, E-mail: devesh.ranjan@me.gatech.edu [Georgia Institute of Technology, George W. Woodruff School of Mechanical Engineering, Atlanta, GA 30332 (United States)

2015-12-15

Highlights: • A code was developed to investigate the various aspects of using cooling tower for S-CO{sub 2} Brayton cycles. • Cooling tower option to reject heat is quantitatively compared to the direct water cooling and dry air cooling options. • Optimum water conditions resulting in minimal plant capital cost per unit power consumption are calculated. - Abstract: A simplified qualitative analysis was performed to investigate the possibility of using counter flow induced draft cooling tower option to reject heat from the supercritical carbon dioxide Brayton cycle for advanced fast reactor (AFR)-100 and advanced burner reactor (ABR)-1000 plants. A code was developed to estimate the tower dimensions, power and water consumption, and to perform economic analysis. The code developed was verified against a vendor provided quotation and is used to understand the effect of ambient air and water conditions on the design of cooling tower. The calculations indicated that there exists optimum water conditions for given ambient air conditions which will result in minimum power consumption, thereby increasing the cycle efficiency. A cost-based optimization technique is used to estimate the optimum water conditions which will improve the overall plant economics. A comparison of different cooling options for the S-CO{sub 2} cycle indicated that the cooling tower option is a much more practical and economical option compared to the dry air cooling or direct water cooling options.

Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

2002-01-01

The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

International Nuclear Information System (INIS)

1976-08-01

This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

Design and Test of Wendelstein 7-X Water-Cooled Divertor Scraper

Energy Technology Data Exchange (ETDEWEB)

Boscary, J. [Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany; Greuner, Henri [Max Planck Institute for Plasma Physics, Garching, Germany; Ehrke, Gunnar [Max Planck Institute of Plasma Physics, Greifswald, Germany; Boeswirth, Bernd [Max Planck Institute for Plasma Physics, Garching, Germany; Wang, Zhongwei [Max Planck Institute for Plasma Physics, Garching, Germany; Clark, Emily [The University of Tennessee, Knoxville; Lumsdaine, Arnold [ORNL; Tretter, Jorg [Max Planck Institute for Plasma Physics, Garching, Germany; Junghanns, Patrick [Max Planck Institute for Plasma Physics, Garching, Germany; Stadler, Reinhold [Max Planck Institute for Plasma Physics, Garching, Germany; McGinnis, William Dean [ORNL; Lore, Jeremy D. [ORNL; Team, W7-X [Max-Planck-Institut fur Plasmaphysik, Griefswald, Germany

2018-04-01

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipes of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.

Adaptation and implementation of the TRACE code for transient analysis in designs lead cooled fast reactors

International Nuclear Information System (INIS)

Lazaro, A.; Ammirabile, L.; Martorell, S.

2015-01-01

Lead-Cooled Fast Reactor (LFR) has been identified as one of promising future reactor concepts in the technology road map of the Generation IVC International Forum (GIF)as well as in the Deployment Strategy of the European Sustainable Nuclear Industrial Initiative (ESNII), both aiming at improved sustainability, enhanced safety, economic competitiveness, and proliferation resistance. This new nuclear reactor concept requires the development of computational tools to be applied in design and safety assessments to confirm improved inherent and passive safety features of this design. One approach to this issue is to modify the current computational codes developed for the simulation of Light Water Reactors towards their applicability for the new designs. This paper reports on the performed modifications of the TRACE system code to make it applicable to LFR safety assessments. The capabilities of the modified code are demonstrated on series of benchmark exercises performed versus other safety analysis codes. (Author)

Potential climate change impacts on water availability and cooling water demand in the Lusatian Lignite Mining Region, Central Europe

Science.gov (United States)

Pohle, Ina; Koch, Hagen; Gädeke, Anne; Grünewald, Uwe; Kaltofen, Michael; Redetzky, Michael

2014-05-01

In the catchments of the rivers Schwarze Elster, Spree and Lusatian Neisse, hydrologic and socioeconomic systems are coupled via a complex water management system in which water users, reservoirs and water transfers are included. Lignite mining and electricity production are major water users in the region: To allow for open pit lignite mining, ground water is depleted and released into the river system while cooling water is used in the thermal power plants. In order to assess potential climate change impacts on water availability in the catchments as well as on the water demand of the thermal power plants, a climate change impact assessment was performed using the hydrological model SWIM and the long term water management model WBalMo. The potential impacts of climate change were considered by using three regional climate change scenarios of the statistical regional climate model STAR assuming a further temperature increase of 0, 2 or 3 K by the year 2050 in the region respectively. Furthermore, scenarios assuming decreasing mining activities in terms of a decreasing groundwater depression cone, lower mining water discharges, and reduced cooling water demand of the thermal power plants are considered. In the standard version of the WBalMo model cooling water demand is considered as static with regard to climate variables. However, changes in the future cooling water demand over time according to the plans of the local mining and power plant operator are considered. In order to account for climate change impacts on the cooling water demand of the thermal power plants, a dynamical approach for calculating water demand was implemented in WBalMo. As this approach is based on air temperature and air humidity, the projected air temperature and air humidity of the climate scenarios at the locations of the power plants are included in the calculation. Due to increasing temperature and decreasing precipitation declining natural and managed discharges, and hence a lower

Thermophysical properties of materials for water cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-06-01

The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA`s International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs.

Thermophysical properties of materials for water cooled reactors

International Nuclear Information System (INIS)

1997-06-01

The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

Comparison of solar panel cooling system by using dc brushless fan and dc water

International Nuclear Information System (INIS)

Irwan, Y M; Leow, W Z; Irwanto, M; M, Fareq; Hassan, S I S; Amelia, A R; Safwati, I

2015-01-01

The purpose of this article is to discuss comparison of solar panel cooling system by using DC brushless fan and DC water pump. Solar photovoltaic (PV) power generation is an interesting technique to reduce non-renewable energy consumption and as a renewable energy. The temperature of PV modules increases when it absorbs solar radiation, causing a decrease in efficiency. A solar cooling system is design, construct and experimentally researched within this work. To make an effort to cool the PV module, Direct Current (DC) brushless fan and DC water pump with inlet/outlet manifold are designed for constant air movement and water flow circulation at the back side and front side of PV module representatively. Temperature sensors were installed on the PV module to detect temperature of PV. PIC microcontroller was used to control the DC brushless fan and water pump for switch ON or OFF depend on the temperature of PV module automatically. The performance with and without cooling system are shown in this experiment. The PV module with DC water pump cooling system increase 3.52%, 36.27%, 38.98%in term of output voltage, output current, output power respectively. It decrease 6.36 °C compare than to PV module without DC water pump cooling system. While DC brushless fan cooling system increase 3.47%, 29.55%, 32.23%in term of output voltage, output current, and output power respectively. It decrease 6.1 °C compare than to PV module without DC brushless fan cooling system. The efficiency of PV module with cooling system was increasing compared to PV module without cooling system; this is because the ambient temperature dropped significantly. The higher efficiency of PV cell, the payback period of the system can be shorted and the lifespan of PV module can also be longer. (paper)

Gas cooled fast reactor 2400 MWTh, status on the conceptual design studies and preliminary safety analysis

International Nuclear Information System (INIS)

Malo, J.Y.; Alpy, N.; Bentivoglio, F.

2009-01-01

The Gas cooled Fast Reactor (GFR) is considered by the French Commissariat a l'Energie Atomique as a promising concept, combining the benefits of fast spectrum and high temperature, using Helium as coolant. A status on the GFR preliminary viability was made at the end of 2007, ending the pre-conceptual design phase. A consistent overall systems arrangement was proposed and a preliminary safety analysis based on operating transient calculations and a simplified PSA had established a global confidence in the feasibility and safety of this baseline concept. Its potential for attractive performances had been pointed out. Compare to the more mature Sodium Fast Reactor technology, no demonstrator has ever been built and the feasibility demonstration will required a longer lead time. The next main project milestone is related to the GFR viability, scheduled in 2012. The current studies consist in revisiting the reactor reference design options as selected at the end of 2007. Most of them are being consolidated by going more in depth in the analysis. Some possible alternatives are assessed. The paper will give a status on the last studies performed on the core design and corresponding neutronics and cycle performance, the Decay Heat Removal strategy and preliminary safety analysis, systems design and balance of plant... This paper is complementary to the Icapp'09 papers 9062 dealing with the Gas cooled Fast Reactor Demonstrator ALLEGRO and 9378 related to GFR transients analysis. (author)

Severe water ingress accident analysis for a Modular High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Zhang Zuoyi; Scherer, Winfried

1997-01-01

This paper analyzes the severe water ingress accidents in the SIEMENS 200MW Modular High Temperature Gas Cooled Reactor (HTR-Module) under the assumption of no active safety protection systems in order to find the safety margin of the current HTR-Module design. A water, steam and helium multi-phase cavity model is originally developed and implemented in the DSNP simulation system. The developed DSNP system is used to simulate the primary circuit of HTR-Module power plant. The comparisons of the models with the TINTE calculations validate the current simulation. After analyzing the effects of blower separation on water droplets, the wall heat storage, etc., it is found that the maximum H 2 O density increase rate in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduces the impulse of the H 2 O in the reactor core. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600degC was not reached in any case. (author)

Is fasting safe? A chart review of adverse events during medically supervised, water-only fasting.

Science.gov (United States)

Finnell, John S; Saul, Bradley C; Goldhamer, Alan C; Myers, Toshia R

2018-02-20

Evidence suggests that fasting, during which only water is consumed, results in potentially health promoting physiological effects. However, peer-reviewed research assessing the safety of water-only fasting is lacking. To address this, we conducted a chart review to describe adverse events (AEs) that occurred during medically supervised, water-only fasting. Electronic charts from patient visits to a residential medical facility from 2006 to 2011 were reviewed. Patients who were at least 21 years of age and water-only fasted for ≥2 consecutive days with a refeeding period equal to half of the fast length were included. Out of 2539 charts, 768 visits met our inclusion and exclusion criteria. AEs were abstracted from chart notes and classified according to CTCAE (v4.03) and MedDRA (v12.1) terminology. Descriptive analysis of AEs is reported. During the protocol period, the highest grade AE (HGAE) in 555 visits was a grade 2 event or lower, in 212 visits it was a grade 3 event, in 1 visit it was a grade 4 event, and there were no grade 5 events. There were 2 (0.002%) visits with a serious adverse event (SAE). The majority of AEs identified were mild (n = 4490, 75%) in nature and known reactions to fasting. To our knowledge, this is the most comprehensive analysis of AEs experienced during medically supervised, water-only fasting conducted to date. Overall, our data indicate that the majority of AEs experienced were mild to moderate and known reactions to fasting. This suggests that the protocol used in this study can be safely implemented in a medical setting with minimal risk of a SAE.

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

Energy Technology Data Exchange (ETDEWEB)

Perret, G.; Pattupara, R. M. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Girardin, G. [Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland); Chawla, R. [Paul Scherrer Inst., 5232 Villigen (Switzerland); Ecole Polytechnique Federale de Lausanne, 1015 Lausanne (Switzerland)

2012-07-01

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

A water-cooling solution for PC-racks of the LHC experiments

CERN Document Server

Vannerem, P

2004-01-01

With ever increasing power consumption and heat dissipation of todays CPUs, cooling of rack-mounted PCs is an issue for the future online farms of the LHC experiments. In order to investigate the viability of a water-cooling solution, a prototype PC-farm rack has been equipped with a commercially available retrofitted heat exchanger. The project has been carried out as a collaboration of the four LHC experiments and the PH-ESS group . This note reports on the results of a series of cooling and power measurements of the prototype rack with configurations of 30 to 48 PCs. The cooling performance of the rack-cooler is found to be adequate; it extracts the heat dissipated by the CPUs efficiently into the cooling water. Hence, the closed PC rack transfers almost no heat into the room. The measurements and the failure tests show that the rack-cooler concept is a viable solution for the future PC farms of the LHC experiments.

Marginal costs of water savings from cooling system retrofits: a case study for Texas power plants

Science.gov (United States)

Loew, Aviva; Jaramillo, Paulina; Zhai, Haibo

2016-10-01

The water demands of power plant cooling systems may strain water supply and make power generation vulnerable to water scarcity. Cooling systems range in their rates of water use, capital investment, and annual costs. Using Texas as a case study, we examined the cost of retrofitting existing coal and natural gas combined-cycle (NGCC) power plants with alternative cooling systems, either wet recirculating towers or air-cooled condensers for dry cooling. We applied a power plant assessment tool to model existing power plants in terms of their key plant attributes and site-specific meteorological conditions and then estimated operation characteristics of retrofitted plants and retrofit costs. We determined the anticipated annual reductions in water withdrawals and the cost-per-gallon of water saved by retrofits in both deterministic and probabilistic forms. The results demonstrate that replacing once-through cooling at coal-fired power plants with wet recirculating towers has the lowest cost per reduced water withdrawals, on average. The average marginal cost of water withdrawal savings for dry-cooling retrofits at coal-fired plants is approximately 0.68 cents per gallon, while the marginal recirculating retrofit cost is 0.008 cents per gallon. For NGCC plants, the average marginal costs of water withdrawal savings for dry-cooling and recirculating towers are 1.78 and 0.037 cents per gallon, respectively.

The Modification of Sodium Polyacrylate Water Solution Cooling Properties by AL2O3

Directory of Open Access Journals (Sweden)

Wojciech Gęstwa

2010-01-01

Based on cooling curves, it can be concluded that for the water solution of sodium polyacrylate with AL2O3 nanoparticles in comparison to water and 10% polymer water solution lower cooling speed is obtained. The cooling medium containing nanoparticles provides lower cooling speed in the smallest surface austenite occurance (500–600 C in the charts of the CTP for most nonalloy structural steels and low-alloy steels. However lower cooling temperature at the beginning of martensitic transformation causes the formation of smaller internal stresses, leading to smaller dimensional changes and hardening deformation. For the quenching media the wetting angle was appointed by the drop-shape method. These studies showed the best wettability of polymer water solution (sodium polyacrylate with the addition of AL2O3 nanoparticles, whose wetting angle was about 65 degrees. Obtaining the smallest wetting angle for the medium containing nanoparticles suggests that the heat transfer to the cooling medium is larger. This allows slower cooling at the same time ensuring its homogeneity. The obtained values of wetting angle confirm the conclusions drawn on the basis of cooling curves and allowus to conclude that in the case of the heat transfer rate it will have a lower value than for water and 10% polymer water solution. In the research on hardened carburized steel samples C10 and 16MnCr5 surface hardness, impact strength and changes in the size of cracks in Navy C-ring sample are examined. On this basis of the obtained results it can be concluded that polymer water solution with nanoparticles allows to obtain a better impact strength at comparable hardness on the surface. Research on the dimensional changes on the basis of the sample of Navy C-ring also shows small dimensional changes for samples carburized and hardened in 10% polymer water solution with the addition of nanoparticles AL2O3. Smaller dimensional changes were obtained for samples of steel 16MnCr5 thanfar C10. The

Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

International Nuclear Information System (INIS)

Badulescu, A.; Groeneveld, D.C.

2000-01-01

Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

Simulation of Two-Phase Natural Circulation Loop for Core Cather Cooling Using Air Water

International Nuclear Information System (INIS)

Revankar, S. T.; Huang, S. F.; Song, K. W.; Rhee, B. W.; Park, R. J.; Song, J. H.

2012-01-01

A closed loop natural circulation system employs thermally induced density gradients in single phase or two-phase liquid form to induce circulation of the working fluid thereby obviating the need for any mechanical moving parts such as pumps and pump controls. This increases the reliability and safety of the cooling system and reduces installation, operation and maintenance costs. That is the reason natural circulation cooling has been considered in advanced reactor core cooling and in engineered safety systems. Natural circulation cooling has been proposed to remove reactor decay heat by external vessel cooling for in-vessel core retention during sever accident scenario. Recently in APR1400 reactor core catcher design natural circulation cooling is proposed to stabilize and cool the corium ejected from the reactor vessel following core melt and breach of reactor vessel. The natural circulation flow is similar to external vessel cooling where water flows through an inclined narrow gap below hot surface and is heated to produce boiling. The two-phase natural circulation enables cooling of the corium pool collected on core catcher. Due to importance of this problem this paper focuses simulation of the two-phase natural circulation through inclined gap using air-water system. Scaling criteria for air-water loop are derived that enable simulation of the flow regimes and natural circulation flow rates in such systems using air-water system

Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

International Nuclear Information System (INIS)

Weaver, K.D.; Sterbentz, J.; Meyer, M.; Lowden, R.; Hoffman, E.; Wei, T.Y.C.

2004-01-01

The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO 2 ) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

Impact of short‐term, repeated water fasting on the weight of mice

Directory of Open Access Journals (Sweden)

Zahra Mishmast

2014-09-01

Full Text Available Introduction: Caloric restriction is a strategy applied for weight loss. Water fasting is a popular way for obesity treatment. However, little is known about the impact of water fasting on weight. Therefore, this study was conducted to investigate the effect of short-term, repeated water fasting on the weight of mice. Methods: In this study, the physiological effect of short-term, repeated water fasting on the weight of female mice was evaluated. At 6 weeks of age, mice were randomly assigned to either repeated-fasting or control group (fed ad libitum. Each group consisted of twenty mice. As the fasting period started, the repeated-fasting group had access to only water; animals had free access to food and water on non-fasting days. Body weight of each group before, during, and after the fasting period was recorded. Results: Body weight of the fasting group significantly decreased, unlike the control group. However, the fasting group gained weight rapidly after being re-fed and became significantly heavier than mice in the control group (P

Custom design of a hanging cooling water power generating system applied to a sensitive cooling water discharge weir in a seaside power plant: A challenging energy scheme

International Nuclear Information System (INIS)

Tian, Chuan Min; Jaffar, Mohd Narzam; Ramji, Harunal Rejan; Abdullah, Mohammad Omar

2015-01-01

In this study, an innovative design of hydro-electricity system was applied to an unconventional site in an attempt to generate electricity from the exhaust cooling water of a coal-fired power plant. Inspired by the idea of micro hydro, present study can be considered new in three aspects: design, resource and site. This system was hung at a cooling water discharge weir, where all sorts of civil work were prohibited and sea water was used as the cooling water. It was designed and fabricated in the university's mechanical workshop and transported to the site for installation. The system was then put into proof run for a three-month period and achieved some success. Due to safety reasons, on-site testing was prohibited by the power plant authority. Hence, most data was acquired from the proof run. The driving system efficiency was tested in the range of 25% and 45% experimentally while modeling results came close to experimental results. Payback period for the system is estimated to be about 4.23 years. Result obtained validates the feasibility of the overall design under the sensitive site application. - Highlights: • Challenging energy scheme via a hanging cooling water power generating system. • Driving system efficiency was tested in the range of 25% and 45%. • Payback period for the system is estimated to be about 4.2 years

Air-cooled LiBr-water absorption chillers for solar air conditioning in extremely hot weathers

International Nuclear Information System (INIS)

Kim, D.S.; Infante Ferreira, C.A.

2009-01-01

A low temperature-driven absorption cycle is theoretically investigated for the development of an air-cooled LiBr-water absorption chiller to be combined with low-cost flat solar collectors for solar air conditioning in hot and dry regions. The cycle works with dilute LiBr-water solutions so that risk of LiBr crystallization is less than for commercially available water-cooled LiBr-water absorption chillers even in extremely hot ambient conditions. Two-phase heat exchangers in the system were modelled taking account of the heat and mass transfer resistances in falling film flows by applying the film theory in thermal and concentration boundary layers. Both directly and indirectly air-cooled chillers were modelled by properly combining component models and boundary conditions in a matrix system and solved with an algebraic equation solver. Simulation results predict that the chillers would deliver chilled water around 7.0 deg. C with a COP of 0.37 from 90 deg. C hot water under 35 deg. C ambient condition. At 50 deg. C ambient temperature, the chillers retained about 36% of their cooling power at 35 deg. C ambient. Compared with the directly air-cooled chiller, the indirectly air-cooled chiller presented a cooling power performance reduction of about 30%

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

International Nuclear Information System (INIS)

Kim, Seokho H.; Berry, Jan

2011-01-01

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclear pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.

Releases from the cooling water system in the Waste Tank Farm

International Nuclear Information System (INIS)

Perkins, W.C.; Lux, C.R.

1991-01-01

On September 12, 1991, a cooling-water header broke in the H-Area Waste Tank farm, at the Savannah River Site, releasing contaminated water down a storm sewer that drains to the creek. A copy of the Occurrence Report is attached. As part of the follow-up on this incident, the NPSR Section was asked by Waste Management Technology to perform a probabilistic analysis of the following cases: (1) A large break in the header combined with a large break in a cooling coil inside a waste tank. (2) A large break in the header combined with a leak in a cooling coil inside a waste tank. (3) A large break in the header combined with a very small leak in a cooling coil inside a waste tank. This report documents the results of the analysis of these cases

Status of advanced technology and design for water cooled reactors: Light water reactors

International Nuclear Information System (INIS)

1988-10-01

Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

Water-cooled beam line components at LAMPF

International Nuclear Information System (INIS)

Grisham, D.L.; Lambert, J.E.

1981-01-01

The beam line components that comprise the main experimental beam at the Clinton P. Anderson Meson Physics Facility (LAMPF) have been operating since February 1976. This paper will define the functions of the primary water-cooled elements, their design evolution, and our operating experience to the present time

Passive vibro-acoustic detection of a sodium-water reaction in a steam generator of a sodium-cooled fast neutrons nuclear reactor by beam forming

International Nuclear Information System (INIS)

Moriot, Jeremy

2013-01-01

This thesis deals with a new method to detect a sodium-water reaction in a steam generator of a fast sodium-cooled nuclear reactor. More precisely, the objective is to detect a micro-leak of water (flow ≤ 1 g/s) in less than 10 seconds by measuring the external shell vibrations of the component. The strong background noise in operation makes impossible the use of a detection system based on a threshold overrun. A beam forming method applied to vibrations measured by a linear array of accelerometers is developed in this thesis to increase the signal-to-noise ratio and to detect and locate the leak in the steam generator. A numerical study is first realized. Two models are developed in order to simulate the signals measured by the accelerometers of the array. The performances of the beam forming are then studied in function of several parameters, such as the source location and frequency, the damping factor, the background noise considered. The first model consists in an infinite plate in contact with a heavy fluid, excited by an acoustic monopole located in this fluid. Analyzing the transverse displacements in the wavenumber domain is useful to establish a criterion to sample correctly the vibration field of the plate. A second model, more representative of the system is also proposed. In this model, an elastic infinite cylindrical shell, filled with a heavy fluid is considered. The finite dimensions in the radial and circumferential directions lead to a modal behavior of the system which impacts the beam forming. Finally, the method is tested on an experimental mock-up which consists in a cylindrical pipe made in stainless steel and filled with water connected to hydraulic circuit. The water flow speed can be controlled by varying the speed of the pump. The acoustic source is generated by a hydro-phone. The performances of the beam forming are studied for different water flow speeds and different amplitude and frequencies of the source. (author) [fr

Adaptive polynomial chaos techniques for uncertainty quantification of a gas cooled fast reactor transient

International Nuclear Information System (INIS)

Perko, Z.; Gilli, L.; Lathouwers, D.; Kloosterman, J. L.

2013-01-01

Uncertainty quantification plays an increasingly important role in the nuclear community, especially with the rise of Best Estimate Plus Uncertainty methodologies. Sensitivity analysis, surrogate models, Monte Carlo sampling and several other techniques can be used to propagate input uncertainties. In recent years however polynomial chaos expansion has become a popular alternative providing high accuracy at affordable computational cost. This paper presents such polynomial chaos (PC) methods using adaptive sparse grids and adaptive basis set construction, together with an application to a Gas Cooled Fast Reactor transient. Comparison is made between a new sparse grid algorithm and the traditionally used technique proposed by Gerstner. An adaptive basis construction method is also introduced and is proved to be advantageous both from an accuracy and a computational point of view. As a demonstration the uncertainty quantification of a 50% loss of flow transient in the GFR2400 Gas Cooled Fast Reactor design was performed using the CATHARE code system. The results are compared to direct Monte Carlo sampling and show the superior convergence and high accuracy of the polynomial chaos expansion. Since PC techniques are easy to implement, they can offer an attractive alternative to traditional techniques for the uncertainty quantification of large scale problems. (authors)

The dismantling of fast reactors: sodium processing

International Nuclear Information System (INIS)

Rodriguez, G.; Berte, M.; Serpante, J.P.

1999-01-01

Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

Minimization of radioactive material deposition in water-cooled nuclear reactors

International Nuclear Information System (INIS)

Ruiz, C.P.; Blaies, D.M.

1988-01-01

This patent describes the method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor which comprises adding zinc ion to water entering the water-bearing vessel. The improvement contains a substantially lower proportion of the /sup 64/Zn isotope than naturally occurring zinc

A comparative study on recycling spent fuels in gas-cooled fast reactors

International Nuclear Information System (INIS)

Choi, Hangbok; Baxter, Alan

2010-01-01

This study evaluates advanced Gas-cooled Fast Reactor (GFR) fuel cycle scenarios which are based on recycling spent nuclear fuel for the sustainability of nuclear energy. A 600 MWth GFR was used for the fuel cycle analysis, and the equilibrium core was searched with different fuel-to-matrix volume ratios such as 70/30 and 60/40. Two fuel cycle scenarios, i.e., a one-tier case combining a Light Water Reactor (LWR) and a GFR, and a two-tier case using an LWR, a Very High Temperature Reactor (VHTR), and a GFR, were evaluated for mass flow and fuel cycle cost, and the results were compared to those of LWR once-through fuel cycle. The mass flow calculations showed that the natural uranium consumption can be reduced by more than 57% and 27% for the one-tier and two-tier cycles, respectively, when compared to the once-through fuel cycle. The transuranics (TRU) which pose a long-term problem in a high-level waste repository, can be significantly reduced in the multiple recycle operation of these options, resulting in more than 110 and 220 times reduction of TRU inventory to be geologically disposed for the one-tier and two-tier fuel cycles, respectively. The fuel cycle costs were estimated to be 9.4 and 8.6 USD/MWh for the one-tier fuel cycle when the GFR fuel-to-matrix volume ratio was 70/30 and 60/40, respectively. However the fuel cycle cost is reduced to 7.3 and 7.1 USD/MWh for the two-tier fuel cycle, which is even smaller than that of the once-through fuel cycle. In conclusion the GFR can provide alternative fuel cycle options to the once-through and other fast reactor fuel cycle options, by increasing the natural uranium utilization and reducing the fuel cycle cost.

Collisions between Whales and Fast Ferries around Korean Waters

Directory of Open Access Journals (Sweden)

Kyung-Jun Song

2013-01-01

Full Text Available Although there is heavy maritime traffic around Korean waters, collisions between whales and fast ferries around Korean waters are nearly unknown. A ship strike that was associated with a minke whale occurred near the southeastern part of Tsushima Island along the sailing route of the fast ferry between Korea and Japan on December 16, 2004. It was associated with a fast ferry that runs at a speed of approximately 46.1 kn (83 km/h between Korea and Japan. This individual was certainly seriously injured or killed by this ship strike because large amounts of skin of this individual were attached to the surface of the fast ferry, and also large amounts of blood of this individual spread out in that area. However, fortunately, serious damage did not occur to the mariners and passengers of the ferry, although many passengers were knocked down to the floor of the fast ferry when the fast ferry collided with the minke whale. In addition, a total of 4 records of possible collisions between whales and fast ferries have occurred on the fast ferry route between Korea and Japan between 2004 and 2007. This study is the first formal report on collisions between whales and fast ferries around Korean waters. Although the effect of ship strikes on the survival of cetaceans distributed around Korean waters is not very high at present compared with that of other threats, such as entanglement in fishing gear, ship strikes can pose a significant potential threat to endangered cetaceans such as western gray whales. Therefore, it is necessary to prepare prevention measures for ship strikes for the conservation of cetaceans around Korean waters in the future.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

Energy Technology Data Exchange (ETDEWEB)

Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

2016-05-15

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Choi, Jin Ha; Kim, Myung Hyun

2016-01-01

The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

A water-cooled 13-kG magnet system

International Nuclear Information System (INIS)

Rossi, J.O.; Goncalves, J.A.N.; Barroso, J.J.; Patire Junior, H.; Spassovsky, I.P.; Castro, P.J.

1993-01-01

The construction, performance, and reliability of a high field magnet system are reported. The magnet is designed to generate a flat top 13 kG magnetic induction required for the operation of a 35 GHz, 100 k W gyrotron under development at INPE. The system comprises three solenoids, located in the gun, cavity, and collector regions, consisting of split pair magnets with the field direction vertical. The magnets are wound from insulated copper tube whose rectangular cross section has 5.0 mm-diameter hole leading the cooling water. On account of the high power (∼ 100 k W) supplied to the cavity coils, it turned out necessary to employ a cooling system which includes hydraulic pump a heat exchanger. The collector and gun magnets operate at lower DC current (∼ 150 A), and, in this case, flowing water provided by wall pipes is far enough to cool down the coils. In addition, a 250 k V A high power AC/DC Nutek converser is used to supply power to the cavity magnet. For the collector and gun magnets, 30 V/600 A DC power supplies are used. (author)

An alternative solution for heavy liquid metal cooled reactors fuel assemblies

Energy Technology Data Exchange (ETDEWEB)

Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

2014-10-15

Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

Some aspects of cooling water discharges and environmental enhancement

International Nuclear Information System (INIS)

Grimaas, U.

1976-01-01

As a consequence of the effects of cooling water discharge on the environment, the siting of nuclear power plants is approached with cautiousness. The pros and cons are discussed of siting near bodies of good quality water or in more densely populated or industrial areas. Properties and effects of thermal discharges are elaborated. The effects of heat on the activity of individual organisms, on the accumulation of organic material, on the mineralization rate of organic matter and on the transport of oxygen all have influences on recipient water bodies. Examples of siting Swedish thermal power stations are described and these indicate some negative effects. However, the results do not repudiate the possibility of good effects from the design of new cooling water intake and discharge systems that would speed up the mineralization of organic matters by addition of heat and oxygen. It is concluded that, when choosing between possible sites, areas should be selected where the available energy of the discharge can be used to improve water quality. (author)

Status of conceptual safety design study of Japanese sodium-cooled fast reactor

International Nuclear Information System (INIS)

Kubo, Shigenobu; Kurisaka, Kenichi; Niwa, Hajime; Shimakawa, Yoshio

2005-01-01

In this paper, the current conceptual safety design and related evaluation of Japanese Sodium-cooled Fast Reactor which is studied in the framework of the Feasibility Study (FS) on commercialized Fast Reactor Cycle Systems in Japan are described. The purpose of the safety design is to establish a feasible safety concept of FBR which aims at a sustainable energy source of the next generations. The safety targets and the safety design principle are set aiming at realizing worldwide acceptability of the safety level. The basic safety design concept, which can meet the safety targets, was formulated taking along with the defense-in-depth philosophy as the basic safety design principle. In order to cope with wide range of energy and resource demands, there are some various designs both of oxide and metal fuel for JSFR. Some analytical results of typical design basis events, design extension conditions and core damage frequency estimation show the feasibility of the safety design concept for them. (author)

Reflections on the introduction of fast breeder reactors in the DeBeNeLux states

International Nuclear Information System (INIS)

Schroeder, R.; Wagner, J.

1975-06-01

This report gives a survey of the impact of introducing sodium-cooled fast breeder reactors in the Federal Republic of Germany and the BeNeLux countries (DeBeNeLux region). The supply situation with respect to electric and thermal energy is studied in particular, together with aspects of economy and environmental impact. The potential and consequences of a breeder economy, the present status and future r+d work are discussed. In addition to sodium-cooled fast breeder reactors with oxide or carbide fuel, alternative solutions are touched: 1) light water and high temperature reactors, 2) helium-cooled fast breeder reactors, 3) geothermal energy, solar energy and fusion energy. (orig.) [de

Thermal and electrical energy yield analysis of a directly water cooled photovoltaic module

Directory of Open Access Journals (Sweden)

Mtunzi Busiso

2016-01-01

Full Text Available Electrical energy of photovoltaic modules drops by 0.5% for each degree increase in temperature. Direct water cooling of photovoltaic modules was found to give improved electrical and thermal yield. A prototype was put in place to analyse the field data for a period of a year. The results showed an initial high performance ratio and electrical power output. The monthly energy saving efficiency of the directly water cooled module was found to be approximately 61%. The solar utilisation of the naturally cooled photovoltaic module was found to be 8.79% and for the directly water cooled module its solar utilisation was 47.93%. Implementation of such systems on households may reduce the load from the utility company, bring about huge savings on electricity bills and help in reducing carbon emissions.

Development of components for the gas-cooled fast breeder reactor program

International Nuclear Information System (INIS)

Dee, J.B.; Macken, T.

1977-01-01

The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

Estimation of the residual bromine concentration after disinfection of cooling water by statistical evaluation.

Science.gov (United States)

Megalopoulos, Fivos A; Ochsenkuehn-Petropoulou, Maria T

2015-01-01

A statistical model based on multiple linear regression is developed, to estimate the bromine residual that can be expected after the bromination of cooling water. Make-up water sampled from a power plant in the Greek territory was used for the creation of the various cooling water matrices under investigation. The amount of bromine fed to the circuit, as well as other important operational parameters such as concentration at the cooling tower, temperature, organic load and contact time are taken as the independent variables. It is found that the highest contribution to the model's predictive ability comes from cooling water's organic load concentration, followed by the amount of bromine fed to the circuit, the water's mean temperature, the duration of the bromination period and finally its conductivity. Comparison of the model results with the experimental data confirms its ability to predict residual bromine given specific bromination conditions.

Heat transfer enhancement in a natural draft dry cooling tower under crosswind operation with heterogeneous water distribution

Energy Technology Data Exchange (ETDEWEB)

Goodarzi, Mohsen; Amooie, Hossein [Bu-Ali Sina Univ., Hamedan (Iran, Islamic Republic of). Dept. of Mechanical Engineering

2016-04-15

Crosswind significantly decreases cooling efficiency of a natural draft dry cooling tower. The possibility of improving cooling efficiency with heterogeneous water distribution within the cooling tower radiators under crosswind condition is analysed. A CFD approach was used to model the flow field and heat transfer phenomena within the cooling tower and airflow surrounding the cooling tower. A mathematical model was developed from various CFD results. Having used a trained Genetic Algorithm with the result of mathematical model, the best water distribution was found among the others. Remodeling the best water distribution with the CFD approach showed that the highest enhancement of the heat transfer compared to the usual uniform water distribution.

Heat transfer enhancement in a natural draft dry cooling tower under crosswind operation with heterogeneous water distribution

International Nuclear Information System (INIS)

Goodarzi, Mohsen; Amooie, Hossein

2016-01-01

Crosswind significantly decreases cooling efficiency of a natural draft dry cooling tower. The possibility of improving cooling efficiency with heterogeneous water distribution within the cooling tower radiators under crosswind condition is analysed. A CFD approach was used to model the flow field and heat transfer phenomena within the cooling tower and airflow surrounding the cooling tower. A mathematical model was developed from various CFD results. Having used a trained Genetic Algorithm with the result of mathematical model, the best water distribution was found among the others. Remodeling the best water distribution with the CFD approach showed that the highest enhancement of the heat transfer compared to the usual uniform water distribution.

Achieving reduced fouling of cooling water exchangers with stainless steel tubes

International Nuclear Information System (INIS)

Iftikhar, A.; Mir, N.

2010-01-01

Good performance of cooling water heat exchangers plays a vital role in the over all energy efficiency of a chemical plant. Heavy fouling on carbon steel tubes of the cooling water exchangers was causing poor performance and frequent cleaning requirement. The carbon steel tubes were replaced with stainless steel tubes. Improved performance was achieved and cleaning frequency reduced. The paper covers the details of study and methodology applied for the above changes along with summary of results. (author)

Energy management techniques: SRP cooling water distribution system

International Nuclear Information System (INIS)

Edenfield, A.B.

1979-10-01

Cooling water for the nuclear reactors at the Savannah River Plant is supplied by a pumping and distribution system that includes about 50 miles of underground pipeline. The energy management program at SRP has thus far achieved a savings of about 5% (186 x 10 9 Btu) of the energy consumed by the electrically powered cooling water pumps; additional savings of about 14% (535 x 10 9 Btu) can be achieved by capital expenditures totaling about $3.7 million. The present cost of electricity for operation of this system is about $25 million per year. A computer model of the system was adapted and field test data were used to normalize the program to accurately represent pipeline physical characteristics. Alternate pumping schemes are analyzed to determine projected energy costs and impact on system safety and reliability

Assessment of the dry process fuel sodium-cooled fast reactors

Energy Technology Data Exchange (ETDEWEB)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed.

Assessment of the dry process fuel sodium-cooled fast reactors

International Nuclear Information System (INIS)

Roh, Gyu Hong; Choi, Hang Bok

2004-04-01

The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

Emergency reactor cooling device

International Nuclear Information System (INIS)

Arakawa, Ken.

1993-01-01

An emergency nuclear reactor cooling device comprises a water reservoir, emergency core cooling water pipelines having one end connected to a water feeding sparger, fire extinguishing facility pipelines, cooling water pressurizing pumps, a diesel driving machine for driving the pumps and a battery. In a water reservoir, cooling water is stored by an amount required for cooling the reactor upon emergency and for fire extinguishing, and fire extinguishing facility pipelines connecting the water reservoir and the fire extinguishing facility are in communication with the emergency core cooling water pipelines connected to the water feeding sparger by system connection pipelines. Pumps are operated by a diesel power generator to introduce cooling water from the reservoir to the emergency core cooling water pipelines. Then, even in a case where AC electric power source is entirely lost and the emergency core cooling system can not be used, the diesel driving machine is operated using an exclusive battery, thereby enabling to inject cooling water from the water reservoir to a reactor pressure vessel and a reactor container by the diesel drive pump. (N.H.)

Sodium pool fire analysis of sodium-cooled fast reactor by calculation

International Nuclear Information System (INIS)

Yu Hong; Xu Mi; Jin Degui

2002-01-01

Theoretical models were established according to the characteristic of sodium pool fire, and the SPOOL code was created independently. Some transient processes in sodium pool fire were modeled, including chemical reaction of sodium and oxygen; sodium combustion heat transfer modes in several kids of media; production, deposition and discharge of sodium aerosol; mass and energy exchange between different media in different ventilating conditions. The important characteristic parameters were calculated, such as pressure and temperature of gas, temperature of building materials, mass concentration of sodium aerosol, and so on. The SPOOL code, which provided available safety analysis tool for sodium pool fire accidents in sodium-cooled fast reactor, was well demonstrated with experimental data

Calculating the evaporated water flow in a wet cooling tower

International Nuclear Information System (INIS)

Grange, J.L.

1994-04-01

On a cooling tower, it is necessary to determine the evaporated water flow in order to estimate the water consumption with a good accuracy according to the atmospheric conditions, and in order to know the characteristics of the plume. The evaporated flow is small compared to the circulating flow. A direct measurement is very inaccurate and cannot be used. Only calculation can give a satisfactory valuation. The two usable theories are the Merkel's one in which there are some simplifying assumptions, and the Poppe's one which is more exact. Both theories are used in the numerical code TEFERI which has been developed and is run by Electricite de France. The results obtained by each method are compared and validated by measurements made in the hot air of a cooling tower. The consequences of each hypothesis of Merkel's theory are discussed. This theory does not give the liquid water content in the plume and it under-estimates the evaporated flow all the lower the ambient temperature is. On the other hand, the Poppe's method agrees very closely with the measurements as well for the evaporated flow than for the liquid water concentration. This method is used to establish the specific consumption curves of the great nuclear plants cooling towers as well as to calculate the emission of liquid water drops in the plumes. (author). 11 refs., 9 figs

Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

International Nuclear Information System (INIS)

Tonny Anak Lanyau; Mohd Fazli Zakaria; Yahya Ismail

2011-01-01

Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H 2 O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

Energy Technology Data Exchange (ETDEWEB)

Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

2012-09-15

Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

Science.gov (United States)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

Gas-Cooled Reactors: the importance of their development

International Nuclear Information System (INIS)

Kasten, P.R.

1978-01-01

Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

The R and D issues necessary to achieve the safety design of commercialized liquid-metal cooled fast reactors

International Nuclear Information System (INIS)

Shoji, Kotake; Koji, Dozaki; Shigenobu, Kubo; Yoshio, Shimakawa; Hajime, Niwa; Masakazu, Ichimiya

2002-01-01

Within the framework of the feasibility study on commercialized fast reactor cycle systems (hereafter described as F/S), the safety design principle is investigated and several kinds of design studies are now in progress. Among the designs for liquid-metal cooled fast reactor (LMR), the advanced loop type sodium cooled fast reactor (FR) is one of the promising candidate as future commercialized LMR. In this paper, the safety related research and development (R and D) issues necessary to achieve the safety design are described along the defence-in-depth principle, taking account of not only the system characteristics of the advanced loop concepts but also design studies and R and D experiences so far. Safety issues related to the hypothetical core disruptive accidents (CDA) are emphasized both from the prevention and mitigation. A re-criticality free core concept with a special fuel assembly is pursued by performing both analytical and experimental efforts, in order to realize the rational design and to establish easy-to-understand safety logic. Sodium related issues are also given to ensure plant availability and to enhance the acceptability to the public. (authors)

Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

International Nuclear Information System (INIS)

Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

2006-01-01

The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

Feasibility study for fast reactor and related fuel cycle. Preliminary studies in 1998

International Nuclear Information System (INIS)

Hayafune, Hiroki; Enuma, Yasuhiro; Kubota, Kenichi; Yoshida, Masashi; Uno, Osamu; Ishikawa, Hiroyasu; Kobayashi, Jun; Umetsu, Youichiro; Ichimiya, Masakazu

1999-10-01

Prior to the feasibility study for fast reactors (FRs) starting from the 1999 fiscal year, planned in the medium and long-term program of JNC, preliminarily studies were performed on 'FR systems except sodium cooled MOX fueled reactors'. Small scale or module type reactors, heavy metal (Pb or Pb-Bi) cooled reactors, gas cooled reactors, light water cooled reactors, and molten salt reactors were studied on the basis of literature. They were evaluated from the viewpoint of the technical possibility (the structure integrity, earthquake resistance, safety, productivity, operability, maintenance repair, difficulty of the development), the long-term targets (market competitiveness as an energy system, utilization of uranium resources, reduction of radioactive waste, security of the non-proliferation), and developmental risk. As the result, the following concepts should be studied for future commercialized FRs. Small scale and module type reactor: Middle-sized reactor with an excellent economical efficiency. Small power reactor with a multipurpose design concept. Gas cooled reactor: CO2 gas cooled reactor, He gas cooled reactor. Heavy metal cooled reactor: Russian type lead cooled reactor. Light water cooled reactor: Light water cooled high converter reactor and super critical pressure light water cooled reactor. Molten salt reactor: Trichloride molten salt reactor which matches the U-Pu cycle. (author)

Microbial speciation and biofouling potential of cooling water used by Ontario Hydro

International Nuclear Information System (INIS)

Sharpe, V.J.

1985-02-01

The cooling water composition and microbial components of biofilms attached to stainless steel wafers submerged in three lake water types were evaluated to determine whether their biofouling potential differed in a predictable manner. The composition of the lake waters was different which affected biofilm composition, where the predominance of specific microbial groups varied between test systems and with time. Some prediction of biofouling potential was possible, and it was concluded that the cooling water in the vicinity of Bruce NGS had the lowest biofouling potential whereas greater biofouling could be expected in the Pickering and Nanticoke stations

Current design efforts for the gas-cooled fast reactor (GFR)

International Nuclear Information System (INIS)

Weaver, K.D.

2005-01-01

Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

International Nuclear Information System (INIS)

Su'ud, Zaki

2008-01-01

In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

Applicability of a desiccant dew-point cooling system independent of external water sources

DEFF Research Database (Denmark)

Bellemo, Lorenzo; Elmegaard, Brian; Kærn, Martin Ryhl

2015-01-01

The applicability of a technical solution for making desiccant cooling systems independent of external water sources is investigated. Water is produced by condensing the desorbed water vapour in a closed regeneration circuit. Desorbed water recovery is applied to a desiccant dew-point cooling...... system, which includes a desiccant wheel and a dew point cooler. The system is simulated during the summer period in the Mediterranean climate of Rome and it results completely independent of external water sources. The seasonal thermal COP drops 8% in comparison to the open regeneration circuit solution...

Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

Energy Technology Data Exchange (ETDEWEB)

K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

2004-09-01

The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

Cooling Water System Monitoring by Means of Mossbauer Spectroscopy

International Nuclear Information System (INIS)

Novakova, A.A.; Pargamotnikas, S.A.; Taseva, V.; Dobbrevsky, I.; Nenov, V.; Bonev, B.

1998-01-01

Mossbauer spectroscopy have been applied to the analysis of corrosion sediments formed on mild steel coupons, which were placed in the different points of the Bourgas Petrochemical Plant Recilculating Cooling Water System. It was shown that the created corrosion products can successfully reflect the ambient water medium pollution to which the coupons were exposed

Performance of materials in the component cooling water systems of pressurized water reactors

International Nuclear Information System (INIS)

Lee, B.S.

1993-01-01

The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

Chemistry control challenges in a supercritical water-cooled reactor

International Nuclear Information System (INIS)

Guzonas, David; Tremaine, Peter; Jay-Gerin, Jean-Paul

2009-01-01

The long-term viability of a supercritical water-cooled reactor (SCWR) will depend on the ability of designers to predict and control water chemistry to minimize corrosion and the transport of corrosion products and radionuclides. Meeting this goal requires an enhanced understanding of water chemistry as the temperature and pressure are raised beyond the critical point. A key aspect of SCWR water chemistry control will be mitigation of the effects of water radiolysis; preliminary studies suggest markedly different behavior than that predicted from simple extrapolations from conventional water-cooled reactor behavior. The commonly used strategy of adding excess hydrogen at concentrations sufficient to suppress the net radiolytic production of primary oxidizing species may not be effective in an SCWR. The behavior of low concentrations of impurities such as transition metal corrosion products, chemistry control agents, anions introduced via make-up water or from ion-exchange resins, and radionuclides (e.g., 60 Co) needs to be understood. The formation of neutral complexes increases with temperature, and can become important under near-critical and supercritical conditions; the most important region is from 300-450 C, where the properties of water change dramatically, and solvent compressibility effects exert a huge influence on solvation. The potential for increased transport and deposition of corrosion products (active and inactive), leading to (a) increased deposition on fuel cladding surfaces, and (b) increased out-of-core radiation fields and worker dose, must be assessed. There are also significant challenges associated with chemistry sampling and monitoring in an SCWR. The typical methods used in current reactor designs (grab samples, on-line monitors at the end of a cooled, depressurized sample line) will be inadequate, and in-situ measurements of key parameters will be required. This paper describes current Canadian activities in SCWR chemistry and chemistry

Water conservation in fasting northern elephant seals (Mirounga angustirostris).

Science.gov (United States)

Lester, Christopher W; Costa, Daniel P

2006-11-01

Prolonged terrestrial fasting is a key element in the life history of elephant seals. While on land seals typically fast without access to fresh water, and thus must maintain positive water balance by reductions in water loss such that they can subsist primarily on metabolic water production (MWP). The terrestrial apnea demonstrated by seals may reduce respiratory evaporative water loss (REWL) to levels that allow seals to make a net gain of water from MWP. We empirically measured REWL in 13 fasting northern elephant seal pups and determined the effects on water conservation of a breathing mode that incorporates a regular pattern of apneas, of > or =1 min in duration, followed by eupneic recovery, compared with a breathing mode with no apneas longer than 20 s and resembling typical breathing patterns in other mammals (normative breathing). Overall REWL fell 41% from 0.075+/-0.013 g min(-1) (mean +/- s.d.) during normative breathing to 0.044+/-0.006 g min(-1) during apneic breathing. The decline in REWL is attributed to a decrease in overall ventilation rate, made possible by a decline in metabolic rate along with an increase in oxygen extraction that would occur during apneic breathing. Data on the range of ambient humidity conditions at the local breeding site were collected and used to bound the range of environmental conditions used in laboratory measurements. Our data showed that the observed variations in ambient humidity had no significant effect on REWL. A combination of apneic breathing and the complex nasal turbinates allows fasting elephant seals to reduce REWL well below the rate of MWP so that they can maintain water balance during the fast.

Chicken meat quality as a function of fasting period and water spray

Directory of Open Access Journals (Sweden)

CM Komiyama

2008-09-01

Full Text Available This study aimed at evaluating the effect of different fasting periods and water spray during lairage on the quality of chicken meat. A number of 300 male Ross broilers were reared up to 42 days of age, and submitted to four pre-slaughter fasting periods (4, 8, 12, and 16 hours and sprayed with water or not during lairage. Deboned breast meat was submitted to the following analysis: pH, color, drip loss, water retention capacity, cooking loss, and shear force. There was a significant effect (p < 0.05 of fasting period on meat luminosity was significantly different, with the highest value obtained for 4-hour fasting, whereas no difference was found among the other fasting periods. Meat pH values were different among fasting periods when birds received water spray, with birds fasted for 4, 8, and 12 hours of fasting presenting lower meat pH values (5.87, 5.87, and 6.04, respectively. The interaction between fasting period and water spray influenced meat drip loss and cooking loss, with birds fasted for 16h and not receiving water spray presenting higher drip loss (4.88 and higher cooking loss (28.24 as compared to the other birds. Fasting period affects meat quality, and very short periods (4h impair meat quality.

Electrochemical filtration for turbidity removal in industrial cooling/process water systems

International Nuclear Information System (INIS)

Kumbhar, A.G.; Venkateswaran, G.

2008-01-01

Water samples of large cooling water reservoirs may look visibly clear and transparent, but still may contain sub-micron size particles at sub-parts-per-million levels. Deposition of these particles on heat exchanger surfaces, reduces the heat transfer efficiency in power industry. In nuclear power plants, additionally it creates radiation exposure problems due to activation of fine metallic turbidity in the reactor core and its subsequent transfer to out-of-core surfaces. Sub-micron filtration creates back high-pressure problem. Zeta filters available commercially are prescribed for separating either positively or negatively charged particles. They are of once-use and throw-type. Precipitation surface modified ion exchangers impart chemical impurities to the system. Thus, sub-micron size and dilute turbidity removal from large volumes of waters such as heat exchanger cooling water in nuclear and power industry poses a problem. Electro deposition of the turbidity causing particles, on porous carbon/graphite felt electrodes, is one of the best suited methods for turbidity removal from large volumes of water due to the filter's high permeability, inertness to the system and regenerability resulting in low waste generation. Initially, active indium turbidity removal from RAPS-1 heavy water moderator system, and microbes removal from heat exchanger cooling lake water of RAPS 1 and 2 were demonstrated with in-house designed and fabricated prototype electrochemical filter (ECF). Subsequently, a larger size, high flow filter was fabricated and deployed for iron turbidity removal from active process waters system of Kaiga Generation Station unit 1 and silica and iron turbidity removal from cooling water pond used for heat exchanger of a high temperature high pressure (HTHP) loop at WSCD, Kalpakkam. The ECF proved its exclusive utility for sub-micron size turbidity removal and microbes removal. ECF maneuverability with potential and current for both positively and

The lead cooled fast reactor benchmark Brest-300: analysis with sensitivity method

International Nuclear Information System (INIS)

Smirnov, V.; Orlov, V.; Mourogov, A.; Lecarpentier, D.; Ivanova, T.

2005-01-01

Lead cooled fast neutrons reactor is one of the most interesting candidates for the development of atomic energy. BREST-300 is a 300 MWe lead cooled fast reactor developed by the NIKIET (Russia) with a deterministic safety approach which aims to exclude reactivity margins greater than the delayed neutron fraction. The development of innovative reactors (lead coolant, nitride fuel...) and fuel cycles with new constraints such as cycle closure or actinide burning, requires new technologies and new nuclear data. In this connection, the tool and neutron data used for the calculational analysis of reactor characteristics requires thorough validation. NIKIET developed a reactor benchmark fitting of design type calculational tools (including neutron data). In the frame of technical exchanges between NIKIET and EDF (France), results of this benchmark calculation concerning the principal parameters of fuel evolution and safety parameters has been inter-compared, in order to estimate the uncertainties and validate the codes for calculations of this new kind of reactors. Different codes and cross-sections data have been used, and sensitivity studies have been performed to understand and quantify the uncertainties sources.The comparison of results shows that the difference on k eff value between ERANOS code with ERALIB1 library and the reference is of the same order of magnitude than the delayed neutron fraction. On the other hand, the discrepancy is more than twice bigger if JEF2.2 library is used with ERANOS. Analysis of discrepancies in calculation results reveals that the main effect is provided by the difference of nuclear data, namely U 238 , Pu 239 fission and capture cross sections and lead inelastic cross sections

Prototype solar heating and cooling systems including potable hot water

Science.gov (United States)

1978-01-01

Progress is reviewed in the development, delivery, and support of two prototype solar heating and cooling systems including potable hot water. The system consisted of the following subsystems: collector, auxiliary heating, potable hot water, storage, control, transport, and government-furnished site data acquisition.

Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

International Nuclear Information System (INIS)

Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

2015-01-01

The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

Energy Performance of Water-based and Air-based Cooling Systems in Plus-energy Housing

DEFF Research Database (Denmark)

Andersen, Mads E.; Schøtt, Jacob; Kazanci, Ongun Berk

2016-01-01

-space, and air-to-water heat pump vs. ground heat exchanger as cooling source) on the system energy performance were investigated while achieving the same thermal indoor conditions. The results show that the water-based floor cooling system performed better than the air-based cooling system in terms of energy...... energy use reductions. The coupling of radiant floor with the ground enables to obtain “free” cooling, although the brine pump power should be kept to a minimum to fully take advantage of this solution. By implementing a ground heat exchanger instead of the heat pump and use the crawl-space air as intake...... air an improvement of 37% was achieved. The cooling demand should be minimized in the design phase as a priority and then the resulting cooling load should be addressed with the most energy efficient cooling strategy. The floor cooling coupled with a ground heat exchanger was shown to be an effective...

Thermal analysis of mass concrete embedded with double-layer staggered heterogeneous cooling water pipes

International Nuclear Information System (INIS)

Yang Jian; Hu Yu; Zuo Zheng; Jin Feng; Li Qingbin

2012-01-01

Removal of hydration heat from mass concrete during construction is important for the quality and safety of concrete structures. In this study, a three-dimensional finite element program for thermal analysis of mass concrete embedded with double-layer staggered heterogeneous cooling water pipes was developed based on the equivalent equation of heat conduction including the effect of cooling water pipes and hydration heat of concrete. The cooling function of the double-layer staggered heterogeneous cooling pipes in a concrete slab was derived from the principle of equivalent cooling. To improve the applicability and precision of the equivalent heat conduction equation under small flow, the cooling function was revised according to its monotonicity and empirical formulas of single-phase forced-convection heat transfer in tube flow. Considering heat hydration of concrete at later age, a double exponential function was proposed to fit the adiabatic temperature rise curve of concrete. Subsequently, the temperature variation of concrete was obtained, and the outlet temperature of cooling water was estimated through the energy conservation principle. Comparing calculated results with actual measured data from a monolith of an arch dam in China, the numerical model was proven to be effective in sufficiently simulating accurate temperature variations of mass concrete. - Highlights: ► Three-dimensional program is developed to model temperature history of mass concrete. ► Massive concrete is embedded with double-layer heterogeneous cooling pipes. ► Double exponential function is proposed to fit the adiabatic temperature rise curve. ► Outlet temperature of cooling water is estimated. ► A comparison is made between the calculated and measured data.