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

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

Directory of Open Access Journals (Sweden)

J. Mao

2018-03-01

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

Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

International Nuclear Information System (INIS)

Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

2002-01-01

A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

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.

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)

Calandria cooling structure in pressure tube reactor

International Nuclear Information System (INIS)

Hyugaji, Takenori; Sasada, Yasuhiro.

1976-01-01

Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

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

Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

International Nuclear Information System (INIS)

Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

2003-01-01

The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

Survey of natural-circulation cooling in U.S. pressurized water reactors

International Nuclear Information System (INIS)

Boyack, B.E.

1985-01-01

Literature describing natural circulation analyses, experiments, and plant operation have been obtained from the Nuclear Regulatory Commission, reactor vendors, utility-sponsored research groups, utilities, national laboratories, and foreign sources. These have been reviewed and significant results and conclusions identified. Three modes of natural-circulation cooling are covered: single phase, two-phase, and reflux condensation. Single-phase natural circulation is amply verified by plant operational data, test data from scaled experimental facilities, and analysis with assessed computer codes. Ample evidence also exists that two-phase natural circulation can successfully cool pressurized water reactors. This mode occurs during certain events such as small-break loss-of-coolant accidents. The data base for reflux condensation is primarily from tests in scaled experimental facilities. There are no plant operational data and only limited assessment of thermal-hydraulic systems codes has been performed. Further work is needed before this mode of natural circulation can be confidently used

ZOCO VI - a computer code to calculate the time- and space-dependent pressure distribution in full pressure containments of water-cooled reactors

International Nuclear Information System (INIS)

Mansfeld, G.

1974-12-01

ZOCO VI is a computer code to investigate the time and space dependent pressure distribution in full pressure containment of water cooled nuclear power reactors following a loss-of-coolant accident, which is caused by the rupture of a main coolant or steam line. ZOCO VI is an improved version of the computer code ZOCO V with enlarged description of condensing events. (orig.) [de

Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information

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)

Pressurized water reactor flow arrangement

International Nuclear Information System (INIS)

Gibbons, J.F.; Knapp, R.W.

1980-01-01

A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

Miniaturized compact water-cooled pitot-pressure probe for flow-field surveys in hypersonic wind tunnels

Science.gov (United States)

Ashby, George C.

1988-01-01

An experimental investigation of the design of pitot probes for flowfield surveys in hypersonic wind tunnels is reported. The results show that a pitot-pressure probe can be miniaturized for minimum interference effects by locating the transducer in the probe support body and water-cooling it so that the pressure-settling time and transducer temperature are compatible with hypersonic tunnel operation and flow conditions. Flowfield surveys around a two-to-one elliptical cone model in a 20-inch Mach 6 wind tunnel using such a probe show that probe interference effects are essentially eliminated.

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)

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

High Pressure Industrial Water Facility

Science.gov (United States)

1992-01-01

In conjunction with Space Shuttle Main Engine testing at Stennis, the Nordberg Water Pumps at the High Pressure Industrial Water Facility provide water for cooling the flame deflectors at the test stands during test firings.

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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.

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. .

Device for the condensation of pressurized steam and its application to the cooling of a nuclear reactor after an incident

International Nuclear Information System (INIS)

Dagard, P.; Couturier, M.

1989-01-01

This document describes an invention which relates to a device for condensation of pressurized water which is at a pressure considerably above atmospheric pressure, such as the steam produced by the steam generator of a pressurized-water nuclear reactor during the cooling of the reactor after an incident. The purpose of the invention is therefore to propose a device for the condensation of steam which is under a pressure which is considerably higher than atmospheric pressure by cooling this circulating steam as a result of contact with a heat-exchange wall which is cooled by water; such a device should be easy to install in a nuclear power plant to ensure passive cooling of the reactor, it should have a very good efficiency because of efficient heat exchangers, and it should require only a limited amount of cooling water in the equipment itself

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)

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)

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

Low-pressure water-cooled inductively coupled plasma torch

Science.gov (United States)

Seliskar, Carl J.; Warner, David K.

1988-12-27

An inductively coupled plasma torch is provided which comprises an inner tube, including a sample injection port to which the sample to be tested is supplied and comprising an enlarged central portion in which the plasma flame is confined; an outer tube surrounding the inner tube and containing water therein for cooling the inner tube, the outer tube including a water inlet port to which water is supplied and a water outlet port spaced from the water inlet port and from which water is removed after flowing through the outer tube; and an r.f. induction coil for inducing the plasma in the gas passing into the tube through the sample injection port. The sample injection port comprises a capillary tube including a reduced diameter orifice, projecting into the lower end of the inner tube. The water inlet is located at the lower end of the outer tube and the r.f. heating coil is disposed around the outer tube above and adjacent to the water inlet.

Experimental testing of cooling by low pressure adsorption in a zeolite

Energy Technology Data Exchange (ETDEWEB)

Redman, C.M.

1985-01-01

A small scale facility was designed, constructed, and utilized to test the use of zeolite adsorption of water vapor to augment chill storage in ice for conventional space cooling. The facility uses solar-derived energy, for the heat source and evaporatively chilled water for the heat sump. The product cooling uses sublimation of ice instead of melting. The ZCAT facility utilizes a heat pumping technique in which a water vapor adsorbent functions as the compressor and condenser. The design was based on use of 13X zeolite as the adsorber because of its high adsorbence at low pressures. However, it has been determined that other materials such as silica gel should give superior performance. While zeolite 13X holds more water in the pressure and temperature ranges of interest, silica gel cycles more water and has less residue water. Both points are very important in the design of an efficient and cost effective system.

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

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

Emergency core cooling system

International Nuclear Information System (INIS)

Arai, Kenji; Oikawa, Hirohide.

1990-01-01

The device according to this invention can ensure cooling water required for emerency core cooling upon emergence such as abnormally, for example, loss of coolant accident, without using dynamic equipments such as a centrifugal pump or large-scaled tank. The device comprises a pressure accumulation tank containing a high pressure nitrogen gas and cooling water inside, a condensate storage tank, a pressure suppression pool and a jet stream pump. In this device there are disposed a pipeline for guiding cooling water in the pressure accumulation tank as a jetting water to a jetting stream pump, a pipeline for guiding cooling water stored in the condensate storage tank and the pressure suppression pool as pumped water to the jetting pump and, further, a pipeline for guiding the discharged water from the jet stream pump which is a mixed stream of pumped water and jetting water into the reactor pressure vessel. In this constitution, a sufficient amount of water ranging from relatively high pressure to low pressure can be supplied into the reactor pressure vessel, without increasing the size of the pressure accumulation tank. (I.S.)

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.

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.

Pressure loadings of Soviet-designed VVER [Water-Cooled, Water-Moderated Energy Reactor] reactor release mitigation structures from large-break LOCAs

International Nuclear Information System (INIS)

Sienicki, J.J.; Horak, W.C.

1989-01-01

Analyses have been carried out of the pressurization of the accident release mitigation structures of Soviet-designed VVER (Water-Cooled, Water-Moderated Energy Reactor) pressurized water reactors following large-break loss-of-coolant accidents. Specific VVER systems for which calculations were performed are the VVER-440 model V230, VVER-440 model V213, and VVER-1000 model V320. Descriptions of the designs of these and other VVER models are contained in the report DOE/NE-0084. The principal objective of the current analyses is to calculate the time dependent pressure loadings inside the accident localization or containment structures immediately following the double-ended guillotine rupture of a primary coolant pipe. In addition, the pressures are compared with the results of calculations of the response of the structures to overpressure. Primary coolant system thermal hydraulic conditions and the fluid conditions at the break location were calculated with the RETRAN-02 Mod2 computer code (Agee, 1984). Pressures and temperatures inside the building accident release mitigation structures were obtained from the PACER (Pressurization Accompanying Coolant Escape from Ruptures) multicompartment containment analysis code developed at Argonne National Laboratory. The analyses were carried out using best estimate models and conditions rather than conservative, bounding-type assumptions. In particular, condensation upon structure and equipment was calculated using correlations based upon analyses of the HDR, Marviken, and Battelle Frankfurt containment loading experiments. The intercompartment flow rates incorporate an effective discharge coefficient and liquid droplet carryover fraction given by expressions of Schwan determined from analyses of the Battelle Frankfurt and Marviken tests. 5 refs., 4 figs

Low pressure cooling seal system for a gas turbine engine

Science.gov (United States)

Marra, John J

2014-04-01

A low pressure cooling system for a turbine engine for directing cooling fluids at low pressure, such as at ambient pressure, through at least one cooling fluid supply channel and into a cooling fluid mixing chamber positioned immediately downstream from a row of turbine blades extending radially outward from a rotor assembly to prevent ingestion of hot gases into internal aspects of the rotor assembly. The low pressure cooling system may also include at least one bleed channel that may extend through the rotor assembly and exhaust cooling fluids into the cooling fluid mixing chamber to seal a gap between rotational turbine blades and a downstream, stationary turbine component. Use of ambient pressure cooling fluids by the low pressure cooling system results in tremendous efficiencies by eliminating the need for pressurized cooling fluids for sealing this gap.

ITER's Tokamak Cooling Water System and the the Use of ASME Codes to Comply with French Regulations of Nuclear Pressure Equipment

International Nuclear Information System (INIS)

Berry, Jan; Ferrada, Juan J.; Curd, Warren; Dell Orco, Giovanni; Barabash, Vladimir; Kim, Seokho H.

2011-01-01

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown. Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements. TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system. U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support

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.

Reactor water spontaneous circulation structure in reactor pressure vessel

International Nuclear Information System (INIS)

Takahashi, Kazumi

1998-01-01

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

Emergency reactor cooling circuit

International Nuclear Information System (INIS)

Araki, Hidefumi; Matsumoto, Tomoyuki; Kataoka, Yoshiyuki.

1994-01-01

Cooling water in a gravitationally dropping water reservoir is injected into a reactor pressure vessel passing through a pipeline upon occurrence of emergency. The pipeline is inclined downwardly having one end thereof being in communication with the pressure vessel. During normal operation, the cooling water in the upper portion of the inclined pipeline is heated by convection heat transfer from the communication portion with the pressure vessel. On the other hand, cooling water present at a position lower than the communication portion forms cooling water lumps. Accordingly, temperature stratification layers are formed in the inclined pipeline. Therefore, temperature rise of water in a vertical pipeline connected to the inclined pipeline is small. With such a constitution, the amount of heat lost from the pressure vessel by way of the water injection pipeline is reduced. Further, there is no worry that cooling water to be injected upon occurrence of emergency is boiled under reduced pressure in the injection pipeline to delay the depressurization of the pressure vessel. (I.N.)

Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

Energy Technology Data Exchange (ETDEWEB)

Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

2006-01-15

This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

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.)

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.

Functional systems of a pressurized water reactor

International Nuclear Information System (INIS)

Heinzel, V.

1982-01-01

The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

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

Reduction of PWR containment pressure after hypothetical accidents by water-cooling of the outer containment surface - annular space spray system

International Nuclear Information System (INIS)

Cremer, J.; Dietrich, D.P.; Roedder, P.

1980-12-01

The consequences of a core melt-out accident in the vicinity of a nuclear power station are determined by the integrity of the safety containment. This can be adversely affected by different events during the course of the core melt-out accident. The most important phenomenon is the contact between the melt and sump water. Due to the evaporation of the sump water, there is a continuous rise in pressure of the safety containment, which finally leads to failure due to excess pressure. In order to reduce the fission product release due to the resulting leakage, one must try to reduce the pressure as quickly as possible. As heat cannot be removed from the steel containment to the environment because of the thick concrete containment, it is best to bypass the insulating effect of the concrete by cooling the steel containment from outside. The aim of this investigation is therefore to work out a technically relatively simple system, which offers the possibility of backfitting, setting to work and repair. Such a system is an annular space spray system, by which the annular space between the concrete and steel containment has water pumped to the level of the dome and evenly sprayed over the top hemisphere. Mobile pumps on fire engines belonging to the fire brigade are sufficient to supply the cooling water and these will be available some hours after the accident occurs. The used spray water without any radioactive components is collected outside the reactor building and/or drained off. (orig./GL) [de

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

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.

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

STUDY OF WATER HAMMERS IN THE FILLING OF THE SYSTEM OF PRESSURE COMPENSATION IN THE WATER-COOLED AND WATER-MODERATED POWER REACTORS

Directory of Open Access Journals (Sweden)

A. V. Korolyev

2017-01-01

Full Text Available The research presented in the article conforms to the severe accident that took place at the Three Mail Island nuclear power plant in the USA. The research is focused on improving the reliability of the pressure compensator that is an important equipment of the primary circuit. In order to simulate such a situation, the stand has been developed to simulate the design of the pressurizer of the PWR-440 reactor, in particular an elliptical shape of the upper lid which has a steam outlet pipe at the top of the construction that creates conditions for occurrence of such water hammers. For the experiments, an installation has been created that makes it possible to measure and record the water hammering that occur when the tanks are filled. Measurement of the amplitude of the water hammering was carried out by a specially developed piezoelectric sensor, and the registration – by a light-beam oscilloscope. The technique of carrying out the experiment is described and the results of an experimental study of the water hammers arising when the vessels are completely filled are presented. Quantitative data were obtained on the amplitudes of the hydraulic impacts depending on the rate of filling of the vessel and the diameter of the outlet, the maximum pressure of the hydraulic shock was 7–9 atm. Comparison of calculated and experimental data has been performed. The allowable discrepancy is explained by the calculated value of the system stiffness coefficient, which did not take into account the presence of welded seams in the tank that imparts the system with additional rigidity. The calculated relationships are obtained, that make it possible to estimate the amplitudes of the water hammers through the acceleration of the water level in front of the outlet from a vessel with an elliptical bottom. The possibility of a water hammer in the pressure compensator is demonstrated by experiment and by theoretical calculations. Based on the experimental data, a

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.)

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)

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)

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.)

A technique to simulate a tube break in a high-pressure gas/cooling water heat exchanger - HTR2008-58161

International Nuclear Information System (INIS)

Antwerpen, H. J. V.; Mulder, E. J.

2008-01-01

The gas cycles of most High Temperature Gas-Cooled Reactors (HTR's) reject heat to water at some stage. In the helium/water heat exchangers of HTR's with direct Brayton cycles, the helium is usually at a much higher pressure than the water. If the pressure boundary between the helium and the water fails inside the heat exchanger. the effect on the rest of the water system has to be established in order to do a proper system design. This can be done most efficiently by using a system simulation code, however, very few system simulation codes has the capability to do gas/liquid interface tracking as required for this problem. This study describes a calculation method with which a gas/liquid heat exchanger tube rupture can be calculated in a simulation code without interface tracking. The course of events after tube rupture is described and appropriate calculation models derived. A mathematical model for a pressure relief valve (PRV) was also created. The calculation models were implemented in the system simulation software Flownex and used to study a tube rupture on a 5000 kPa helium/water heat exchanger. The assembled calculation network solved stable and within reasonable time. The simulation provided insight into the course of events following the tube break. It was shown that the acceleration of water out of the helium cooler, by choked-flow helium, caused the main pressure pulses during the event. The maximum pressure in the water loop occurs on the opposite side of the helium cooler due to constructive interference of the initial pressure wave with itself. It was also shown that by changing only pipe lengths, the system could become prone to severe oscillations after a tube rupture event. (authors)

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Procedure for operating a heavy water cooled power reactor

International Nuclear Information System (INIS)

Rau, P.; Kumpf, H.

1981-01-01

Nuclear reactors cooled by heavy water usually have equipment for fuel element exchange during operation, with the primary circuit remaining contained. This fuel element exchange equipment is expensive and complicated in many respects. According to the invention, the heavy water is therefore replaced by light water after a certain time of operation in such way that light water is led in and heavy water is led off. After the replacement, at least a quarter of the fuel elements of the reactor core is exchanged with the reactor pressure vessel being open. Then the light water serving as a shielding is replaced by heavy water, with the reactor pressure vessel being closed. The invention is of interest particularly for high-conversion reactors. (orig.) [de

Measurement of digital blood pressure after local cooling

DEFF Research Database (Denmark)

Nielsen, S L; Lassen, N A

1977-01-01

A double-inlet plastic cuff was designed for local cooling and systolic blood pressure measurement on the middle phalanx of the fingers. With a tourniquet on the proximal phalanx of one finger, cooling for 5 min made the digital artery temperature equal the skin temperature. The difference between...... the systolic pressure in a control finger and in the cooled finger give the reopening pressure in the digital arteries. At 30, 25, 20, 15, and 10 degrees C, respectively the percent decrease of the finger pressure was 0.2 (0.2), 1.5 (2.5), 8.5 (3.7), 11.4 (3.4), and 15.3 (3.1) in normal young women...

Effect of pressure on the vacuum cooling of iceberg lettuce

Energy Technology Data Exchange (ETDEWEB)

Ozturk, Hande Mutlu [Pamukkale University, Food Engineering Department, Faculty of Engineering, Denizli (Turkey); Ozturk, Harun Kemal [Pamukkale University, Mechanical Engineering Department, Faculty of Engineering, 20070 Kinikli, Denizli (Turkey)

2009-05-15

Vacuum cooling is known as a rapid evaporative cooling technique for any porous product which has free water. The aim of this paper is to apply vacuum cooling technique to the cooling of the iceberg lettuce and show the pressure effect on the cooling time and temperature decrease. The results of vacuum cooling are also compared with conventional cooling (cooling in refrigerator) for different temperatures. Vacuum cooling of iceberg lettuce at 0.7 kPa is about 13 times faster than conventional cooling of iceberg lettuce at 6 C. It has been also found that it is not possible to decrease the iceberg lettuce temperature below 10 C if vacuum cooling method is used and vacuum pressure is set to 1.5 kPa. (author) [French] Le refroidissement sous vide est connu comme une technique evaporative rapide refroidissant pour n'importe quel produit poreux qui a de l'eau libre. Le but de ce papier est d'appliquer le refroidissement sous vide pour le refroidissement de la laitue et examiner l'effet de la pression sur le temps de refroidissement et la diminution de temperature. Les resultats de refroidissement sous vide sont aussi compares avec le refroidissement conventionnel (refroidissement dans le refrigerateur) pour les differentes temperatures. Le refroidissement a vide de laitue a 0.7 kPa est environ 13 fois plus vite que le refroidissement conventionnel de laitue croquante a 6 C. Il a ete aussi constate qu'il n'est pas possible de diminuer la temperature de laitue ci-dessous 10 C si le refroidissement sous vide est utilise comme methode et la pression a vide est montree a 1.5 kPa. (orig.)

Emergency core cooling device

International Nuclear Information System (INIS)

Suzaki, Kiyoshi; Inoue, Akihiro.

1979-01-01

Purpose: To improve core cooling effect by making the operation region for a plurality of water injection pumps more broader. Constitution: An emergency reactor core cooling device actuated upon failure of recycling pipe ways is adapted to be fed with cooling water through a thermal sleeve by way of a plurality of water injection pump from pool water in a condensate storage tank and a pressure suppression chamber as water feed source. Exhaust pipes and suction pipes of each of the pumps are connected by way of switching valves and the valves are switched so that the pumps are set to a series operation if the pressure in the pressure vessel is high and the pumps are set to a parallel operation if the pressure in the pressure vessel is low. (Furukawa, Y.)

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...

Probabilistic fracture mechanics analysis of boiling water reactor vessel for cool-down and low temperature over-pressurization transients

Energy Technology Data Exchange (ETDEWEB)

Park, Jeong Soon; Choi, Young Hwan; Jhung, Myung Jo [Safety Research Division, Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-04-15

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

Study on primary coolant system depressurization effect factor in pressurized water reactor

International Nuclear Information System (INIS)

Ji Duan; Cao Xuewu

2006-01-01

The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

Emergency cooling device for reactors

International Nuclear Information System (INIS)

Inoue, Hisamichi; Naito, Masanori; Sato, Chikara; Chino, Koichi.

1975-01-01

Object: To pour high pressure cooling water into a core, when coolant is lost in a boiling water reactor, thereby restraining the rise of fuel cladding. Structure: A control rod guiding pipe, which is moved up and down by a control rod, is mounted on the bottom of a pressure vessel, the control rod guiding pipe being communicated with a high pressure cooling water tank positioned externally of the pressure vessel, and a differential in pressure between the pressure vessel and the aforesaid tank is detected when trouble of coolant loss occurs, and the high pressure cooling water within the tank is poured into the core through the control rod guiding pipe to restrain the rise of fuel cladding. (Kamimura, M.)

Safety research on fusion DEMO in Japan: Toward development of safety strategy of a water-cooled DEMO

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Makoto, E-mail: nakamura.makoto@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Tobita, Kenji; Someya, Youji; Utoh, Hiroyasu; Sakamoto, Yoshiteru [Japan Atomic Energy Agency, Rokkasho 039-3212, Aomori (Japan); Gulden, Werner [Fusion for Energy, Garching D-85748 (Germany)

2016-11-01

Highlights: • This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. • We report analyses of two transients: (i) complete loss of decay heat removal and (ii) major ex-VV LOCA. • The MELCOR analysis has clarified the temperature histories of the DEMO components in complete loss of decay heat removal. • A strategy to reduce the pressure load to the final barrier confining radioactive materials is proposed against the major ex-VV LOCA. - Abstract: This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier.

Electrochemistry of Water-Cooled Nuclear Reactors

International Nuclear Information System (INIS)

Dgiby Macdonald; Mirna Urquidi-Macdonald; John Mahaffy; Amit Jain Han Sang Kim; Vishisht Gupta; Jonathan Pitt

2006-01-01

This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or ''radiation fields'' around the primary loop and the vessel, as a function of the operating parameters and the water chemistry

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.

Performance Optimization of the Water Cooling System for Resonance Frequency Control of the PEFP DTL

International Nuclear Information System (INIS)

Kim, K. Y.; Kim, H. K.; Kim, H. S.; Yoon, J. C.; Sohn, Y. K.; Kweon, S. J.; Park, J.; Kim, K. S.

2010-03-01

The objective of in this research project is prototype cooling water skid of separated closed loop in order to supply and withdraw low conductivity deionized water in drift tube of drift tube linac as core components of proton accelerates. This report is dealt with design specification of J-PARC 400 MeV Linac cooling water system, PEFP DTL cooling system, specification of RCCS21-24, RCCS101 with pump, loss coefficient for DTL2 modeling, pressure drop with flow rate of heat exchanger.

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.

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

Film cooling adiabatic effectiveness measurements of pressure side trailing edge cooling configurations

Directory of Open Access Journals (Sweden)

R. Becchi

2015-12-01

Full Text Available Nowadays total inlet temperature of gas turbine is far above the permissible metal temperature; as a consequence, advanced cooling techniques must be applied to protect from thermal stresses, oxidation and corrosion the components located in the high pressure stages, such as the blade trailing edge. A suitable design of the cooling system for the trailing edge has to cope with geometric constraints and aerodynamic demands; state-of-the-art of cooling concepts often use film cooling on blade pressure side: the air taken from last compressor stages is ejected through discrete holes or slots to provide a cold layer between hot mainstream and the blade surface. With the goal of ensuring a satisfactory lifetime of blades, the design of efficient trailing edge film cooling schemes and, moreover, the possibility to check carefully their behavior, are hence necessary to guarantee an appropriate metal temperature distribution. For this purpose an experimental survey was carried out to investigate the film covering performance of different pressure side trailing edge cooling systems for turbine blades. The experimental test section consists of a scaled-up trailing edge model installed in an open loop suction type test rig. Measurements of adiabatic effectiveness distributions were carried out on three trailing edge cooling system configurations. The baseline geometry is composed by inclined slots separated by elongated pedestals; the second geometry shares the same cutback configuration, with an additional row of circular film cooling holes located upstream; the third model is equipped with three rows of in-line film cooling holes. Experiments have been performed at nearly ambient conditions imposing several blowing ratio values and using carbon dioxide as coolant in order to reproduce a density ratio close to the engine conditions (DR=1.52. To extend the validity of the survey a comparison between adiabatic effectiveness measurements and a prediction by

Emergency core cooling system

International Nuclear Information System (INIS)

Ando, Masaki.

1987-01-01

Purpose: To actuate an automatic pressure down system (ADS) and a low pressure emergency core cooling system (ECCS) upon water level reduction of a nuclear reactor other than loss of coolant accidents (LOCA). Constitution: ADS in a BWR type reactor is disposed for reducing the pressure in a reactor container thereby enabling coolant injection from a low pressure ECCS upon LOCA. That is, ADS has been actuated by AND signal for a reactor water level low signal and a dry well pressure high signal. In the present invention, ADS can be actuated further also by AND signal of the reactor water level low signal, the high pressure ECCS and not-operation signal of reactor isolation cooling system. In such an emergency core cooling system thus constituted, ADS operates in the same manner as usual upon LOCA and, further, ADS is operated also upon loss of feedwater accident in the reactor pressure vessel in the case where there is a necessity for actuating the low pressure ECCS, although other high pressure ECCS and reactor isolation cooling system are not operated. Accordingly, it is possible to improve the reliability upon reactor core accident and mitigate the operator burden. (Horiuchi, T.)

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...

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

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.

Cooling device for reactor container

International Nuclear Information System (INIS)

Arai, Kenji.

1996-01-01

Upon assembling a static container cooling system to an emergency reactor core cooling system using dynamic pumps in a power plant, the present invention provides a cooling device of lowered center of gravity and having a good cooling effect by lowering the position of a cooling water pool of the static container cooling system. Namely, the emergency reactor core cooling system injects water to the inside of a pressure vessel using emergency cooling water stored in a suppression pool as at least one water source upon loss of reactor coolant accident. In addition, a cooling water pool incorporating a heat exchanger is disposed at the circumference of the suppression pool at the outside of the container. A dry well and the heat exchanger are connected by way of steam supply pipes, and the heat exchanger is connected with the suppression pool by way of a gas exhaustion pipe and a condensate returning pipeline. With such a constitution, the position of the heat exchanger is made higher than an ordinary water level of the suppression pool. As a result, the emergency cooling water of the suppression pool water is injected to the pressure vessel by the operation of the reactor cooling pumps upon loss of coolant accident to cool the reactor core. (I.S.)

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

Experimental study of the application of two trickle media for inlet air pre-cooling of natural draft dry cooling towers

International Nuclear Information System (INIS)

He, Suoying; Guan, Zhiqiang; Gurgenci, Hal; Hooman, Kamel; Lu, Yuanshen; Alkhedhair, Abdullah M.

2015-01-01

Highlights: • Two trickle media were experimentally studied in a low-speed wind tunnel. • Correlations for cooling efficiency and pressure drop were developed. • Both trickle media were proven to have relatively low pressure drops. • Both trickle media had severe water entrainment at large air velocities. - Abstract: This paper is part two of a broader investigation into pre-cooling the air that enters natural draft dry cooling towers. Evaporative cooling of air is to some extent different from evaporative cooling of water. Two trickle media (Trickle125 and Trickle100) originally designed for evaporative cooling of water were studied in an open-circuit wind tunnel for evaporative cooling of air. Three medium thicknesses (200, 300 and 450 mm) and two water flow rates (10 and 5 l/min per m 2 horizontally exposed surface area) were used in the tests. The air velocities ranged from 0.5 to 3.0 m/s. The cooling efficiency and the pressure drop of the two media were curve fitted to yield a set of correlations. The pressure drop ranges for Trickle125 and Trickle100 were 0.7–50 Pa and 0.6–41.6 Pa, respectively. The cooling efficiencies of Trickle125 and Trickle100 fell within 15.7–55.1% and 11–44.4%, respectively. Generally, media with large effective surfaces provide high cooling efficiencies and high pressure drops; there is a trade-off between cooling efficiency and pressure drop when selecting a particular medium for a specific application. The water entrainment off the media was detected with water-sensitive papers, and both media had severe water entrainment at large air velocities

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

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

Mathematical Modeling – The Impact of Cooling Water Temperature Upsurge on Combined Cycle Power Plant Performance and Operation

Science.gov (United States)

Indra Siswantara, Ahmad; Pujowidodo, Hariyotejo; Darius, Asyari; Ramdlan Gunadi, Gun Gun

2018-03-01

This paper presents the mathematical modeling analysis on cooling system in a combined cycle power plant. The objective of this study is to get the impact of cooling water upsurge on plant performance and operation, using Engineering Equation Solver (EES™) tools. Power plant installed with total power capacity of block#1 is 505.95 MWe and block#2 is 720.8 MWe, where sea water consumed as cooling media at two unit condensers. Basic principle of analysis is heat balance calculation from steam turbine and condenser, concern to vacuum condition and heat rate values. Based on the result shown graphically, there were impact the upsurge of cooling water to increase plant heat rate and vacuum pressure in condenser so ensued decreasing plant efficiency and causing possibility steam turbine trip as back pressure raised from condenser.

Reactor core cooling device for nuclear power plant

International Nuclear Information System (INIS)

Tsuda, Masahiko.

1992-01-01

The present invention concerns a reactor core cooling facility upon rupture of pipelines in a BWR type nuclear power plant. That is, when rupture of pipelines should occur in the reactor container, an releasing safety valve operates instantly and then a depressurization valve operates to depressurize the inside of a reactor pressure vessel. Further, an injection valve of cooling water injection pipelines is opened and cooling water is injected to cool the reactor core from the time when the pressure is lowered to a level capable of injecting water to the pressure vessel by the static water head of a pool water as a water source. Further, steams released from the pressure vessel and steams in the pressure vessel are condensed in a high pressure/low pressure emergency condensation device and the inside of the reactor container is depressurized and cooled. When the reactor is isolated, since the steams in the pressure vessel are condensed in the state that the steam supply valve and the return valve of a steam supply pipelines are opened and a vent valve is closed, the reactor can be maintained safely. (I.S.)

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...

The effect of pressurizer-water-level on the low frequency component of the pressure spectrum in a PWR

International Nuclear Information System (INIS)

Por, G.; Izsak, E.; Valko, J.

1984-09-01

The pressure fluctuations were measured in the cooling system of the Paks-1 reactor. A shift of the peak was detected in low frequency component of the pressure fluctuation spectrum which is due to the fluctuations of water level in the pressurizer. Using the model of Katona and Nagy (1983), the eigenfrequencies, the magnitude of the shift and the sensitivity to the pressurizer water level were reproduced in good accordance with the experimental data. (D.Gy.)

A simpler, safer, higher performance cooling system arrangement for water cooled divertors

International Nuclear Information System (INIS)

Carelli, M.D.; Kothmann, R.E.; Green, L.; Zhan, N.J.; Stefani, F.; Roidt, R.M.

1994-01-01

A cooling system arrangement is presented which is specifically designed for high heat flux water cooled divertors. The motivation behind the proposed open-quotes unichannelclose quotes configuration is to provide maximum safety; this design eliminates flow instabilities liable to occur in parallel channel designs, it eliminates total blockage, it promotes cross flow to counteract the effects of partial blockage and/or local hot spots, and it is much more tolerant to the effects of debonding between the beryllium armor and the copper substrate. Added degrees of freedom allow optimization of the design, including the possibility of operating at very high heat transfer coefficients associated with nucleate boiling, while at the same time providing ample margin against departure from nucleate boiling. Projected pressure drop, pumping power, and maximum operating temperatures are lower than for conventional parallel channel designs

Thermodynamic analysis of a new combined cooling and power system using ammonia–water mixture

International Nuclear Information System (INIS)

Wang, Jiangfeng; Wang, Jianyong; Zhao, Pan; Dai, Yiping

2016-01-01

Highlights: • A new combined cooling and power system is proposed. • Exergy destruction analysis is used to identify irreversibility of components in system. • Thermodynamic parameter analysis is performed for system. - Abstract: In order to achieve both power and cooling supply for users, a new combined cooling and power system using ammonia–water mixture is proposed to utilizing low grade heat sources, such as industrial waste heat, solar energy and geothermal energy. The proposed system combines a Kalina cycle and an ammonia–water absorption refrigeration cycle, in which the ammonia–water turbine exhaust is delivered to a separator to extract purer ammonia vapor. The purer ammonia vapor enters an evaporator to generate refrigeration output after being condensed and throttled. Mathematical models are established to simulate the combined system under steady-state conditions. Exergy destruction analysis is conducted to display the exergy destruction distribution in the system qualitatively and the results show that the major exergy destruction occurs in the heat exchangers. Finally a thermodynamic sensitivity analysis is performed and reveals that with an increase in the pressure of separator I or the ammonia mass fraction of basic solution, thermal efficiency and exergy efficiency of the system increase, whereas with an increase in the temperature of separator I, the ammonia–water turbine back pressure or the condenser II pressure, thermal efficiency and exergy efficiency of the system drop.

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.

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

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)

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

Method of avoiding hazards resulting from accidents in water-cooled nuclear reactors

International Nuclear Information System (INIS)

Dorner, S.; Schretzmann, K.; Schumacher, G.

1984-01-01

In water-cooled reactors, e.g. BWRs and PWRs, elemental hydrogen is released by hydrolysis (in-leakage). In case of an accident in these reactors or at emergency cooling of e.g., a gas-cooled reactor with water additional hydrogen is produced by chemical reactions of the water with the cladding material. In order to prevent hydrogen pressurizing and the formation of a detonating gas mixture, dry powder containers are provided for in the endangered compartments of the reactor. In case of danger powdered CuO, MnO 2 , Fe 2 O 3 , or CdO, the oxygen content of which recombines with the hydrogen, is ejected from them. In addition, an extinguishing substance with an anticatalytic resp. inhibition effect and/or an inert gas of the group N 2 , He, Ar, CO 2 may be admixed to the powder resp. powder mixture. (orig./PW)

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank

Science.gov (United States)

Corletti, Michael M.; Lau, Louis K.; Schulz, Terry L.

1993-01-01

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

A computer program for accident calculations of a standard pressurized water reactor

International Nuclear Information System (INIS)

Keutner, H.

1979-01-01

In this computer program the dynamic of a standard pressurized water reactor should be realized by both circulation loops with all important components. All important phenomena are taken into consideration, which appear for calculation of disturbances in order to state a realistic process for some minutes after a disturbance or a desired change of condition. In order to optimize the computer time simplifications are introduced in the statement of a differential-algebraic equalization system such that all important effects are taken into consideration. The model analysis starts from the heat production of the fuel rod via cladding material to the cooling medium water and considers the delay time from the core to the steam generator. Alternations of the cooling medium pressure as well as the different temperatures in the primary loop influence the pressuring system - the pressurizer - which is realized by a water and a steam zone with saturated and superheated steam respectively saturated and undercooled water with injection, heating and blow-down devices. The bilance of the steam generator to the secondary loop realizes the process engineering devices. Thereby the control regulation of the steam pressure and the reactor performance is realized. (orig.) [de

Design and analysis on super-critical water cooled power reactors

International Nuclear Information System (INIS)

Ishiwatari, Yuki

2005-01-01

The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

Simulation of the solidification in a channel of a water-cooled glass flow

Directory of Open Access Journals (Sweden)

G. E. Ovando Chacon

2014-12-01

Full Text Available A computer simulation study of a laminar steady-state glass flow that exits from a channel cooled with water is reported. The simulations are carried out in a two-dimensional, Cartesian channel with a backward-facing step for three different angles of the step and different glass outflow velocities. We studied the interaction of the fluid dynamics, phase change and thermal behavior of the glass flow due to the heat that transfers to the cooling water through the wall of the channel. The temperature, streamline, phase change and pressure fields are obtained and analyzed for the glass flow. Moreover, the temperature increments of the cooling water are characterized. It is shown that, by reducing the glass outflow velocity, the solidification is enhanced; meanwhile, an increase of the step angle also improves the solidification of the glass flow.

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

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.

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

Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

International Nuclear Information System (INIS)

Hareux, F.; Roche, R.; Vrillon, B.

1964-01-01

Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

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

Chooz A: a model for the dismantling of water-cooled reactors

International Nuclear Information System (INIS)

Anon.

2017-01-01

The specificity of Chooz-A, the first French pressurized water reactor (PWR), is that the reactor and its major components (pumps, exchangers and cooling circuits) are installed in 2 caves dug out in a hill slope. Chooz-A was operating from 1967 to 1991, in 1993 the fuel was removed and in 2007 EDF received the authorization to dismantle the reactor. In 2012, EDF completed the dismantling of the cave containing the elements of the cooling circuit, a cornerstone was the removing of the four 14 m high steam generators. The dismantling of the pressure vessel began in march 2017, it is the same tools and the same processes that were used for the dismantling of the pressure vessel of the Zorita plant (Spain) in 2016. The end of the Chooz-A dismantling is expected in 2022. The feedback experience will help to standardize practices for the French fleet of PWRs. (A.C.)

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.)

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

Comparison of Cooling Different Parts in a High Pressure Ratio Centrifugal Compressor

Directory of Open Access Journals (Sweden)

S. Mostafa Moosania

2016-12-01

Full Text Available Cooling in a centrifugal compressor can improve the performance and reduce the impeller temperature. In a centrifugal compressor, external walls can be cool down, which is known as the shell cooling. This method avoids undesirable effects induced by other cooling methods. Cooling can be applied on different external walls, such as the shroud, diffuser or the back plate. This paper focuses on seeking the most effective cooling place to increase the performance and reduce the impeller temperature. It is found that shroud cooling improves the compressor performance the most. Shroud cooling with 2400 W of cooling power increases the pressure ratio by 4.6% and efficiency by 1.49%. Each 500 W increase in the shroud cooling power, increases the efficiency by 0.3%. Diffuser cooling and back plate cooling have an identical effect on the polytropic efficiency. However, back plate cooling increases the pressure ratio more than diffuser cooling. Furthermore, only back plate cooling reduces the impeller temperature, and with 2400 W of cooling power, the impeller temperature reduces by 45 K.

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

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

Emergency core cooling system

International Nuclear Information System (INIS)

Kato, Ken.

1989-01-01

In PWR type reactors, a cooling water spray portion of emergency core cooling pipelines incorporated into pipelines on high temperature side is protruded to the inside of an upper plenum. Upon rupture of primary pipelines, pressure in a pressure vessel is abruptly reduced to generate a great amount of steams in the reactor core, which are discharged at a high flow rate into the primary pipelines on high temperature side. However, since the inside of the upper plenum has a larger area and the steam flow is slow, as compared with that of the pipelines on the high temperature side, ECCS water can surely be supplied into the reactor core to promote the re-flooding of the reactor core and effectively cool the reactor. Since the nuclear reactor can effectively be cooled to enable the promotion of pressure reduction and effective supply of coolants during the period of pressure reduction upon LOCA, the capacity of the pressure accumulation vessel can be decreased. Further, the re-flooding time for the reactor is shortened to provide an effect contributing to the improvement of the safety and the reduction of the cost. (N.H.)

Emergency cooling system for a gas-cooled nuclear reactor

International Nuclear Information System (INIS)

Cook, R.K.; Burylo, P.S.

1975-01-01

The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

Experimental Study and Engineering Practice of Pressured Water Coupling Blasting

Directory of Open Access Journals (Sweden)

J. X. Yang

2017-01-01

Full Text Available Overburden strata movement in large space stope is the major reason that induces the appearance of strong mining pressure. Presplitting blasting for hard coal rocks is crucial for the prevention and control of strong pressure in stope. In this study, pressured water coupling blasting technique was proposed. The process and effect of blasting were analyzed by orthogonal test and field practice. Results showed that the presence of pressure-bearing water and explosive cartridges in the drill are the main influence factors of the blasting effect of cement test block. The high load-transmitting performance of pore water and energy accumulation in explosive cartridges were analyzed. Noxious substances produced during the blasting process were properly controlled because of the moistening, cooling, and diluting effect of pore water. Not only the goal of safe and static rock fragmentation by high-explosive detonation but also a combination of superdynamic blast loading and static loading effect of the pressured water was achieved. Then the practice of blasting control of hard coal rocks in Datong coal mine was analyzed to determine reasonable parameters of pressured water coupling blasting. A good presplitting blasting control effect was achieved for the hard coal rocks.

Comparison of wet and dry heat transfer and pressure drop tests of smooth and rough corrugated PVC packing in cooling towers

International Nuclear Information System (INIS)

Goshayeshi, H.R.; Missenden, J.F.

1998-01-01

This paper presents the results of an experimental investigation of the performance of a cooling tower with PVC packing. The following were examined; the effect of surface roughness, the effect of the angle of roughness and the effect of packing spacing. The investigation was divided into two parts: comparison of film heat transfer with air pressure drop, without water circulation and comparison of enthalpy change and pressure drop in the model cooling tower, with circulation of water. Seven commercial packing were investigated, covering a size range of 1.1< P/D<1.70 and 1≤p/e≤5 and a discussion of the dimensionless correlation resulting is given

Cooling system for auxiliary reactor component

International Nuclear Information System (INIS)

Fujihira, Tomoko.

1991-01-01

A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

A passive emergency heat sink for water cooled reactors with particular application to CANDU reg-sign reactors

International Nuclear Information System (INIS)

Spinks, N.J.

1996-01-01

Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU reg-sign moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners

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

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

2014-08-01

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-08-15

The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

Cooling methods of station blackout scenario for LWR plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

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

Pressure suppression facility for reactor container

International Nuclear Information System (INIS)

Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

1993-01-01

In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

Fast reactor cooled by supercritical light water

Energy Technology Data Exchange (ETDEWEB)

Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

2001-09-01

This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

Directory of Open Access Journals (Sweden)

Soo-Yong Park

2015-10-01

Full Text Available Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

International Nuclear Information System (INIS)

Park, Soo Yong; Ahn, Kwang Il

2015-01-01

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO

Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

2009-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

Pressure Separators for District Cooling; Tryckvaexlare foer fjaerrkyla - Teknik och funktion

Energy Technology Data Exchange (ETDEWEB)

Gustafson, Bror-Arne [Fludex AB, Goeteborg (SE)] [and others

2003-07-01

About 10 pressure separators run today in commercial operation in district heating networks. Undoubtedly, the pressure separator has become a new tool for more efficient operation of district heating systems. The pressure separator makes it possible to keep different parts of the pipe network at different pressure levels without the unavoidable temperature losses of heat exchangers. The objectives of this project are to find the answers of two questions. The first question is: Will the pressure separator function in district cooling systems if it is designed in the same way as for district heating? The only difference should then be the temperatures of operation. The second question is: Is there a modified design that will perform better in district cooling systems? To find the answers of the two questions, a test rig was built in Rosenlund Power Station in Goeteborg. Also computer simulations were carried out to clarify actual phenomena. The answer to the first question is: Yes/no. Measurements show that the original design of the pressure separator can be made working at temperatures of operation typical for district cooling. It will, however, be very sensitive and is not recommended for practical applications. The answer to the second question is: Yes, there is. This makes the details of the original design less interesting. The modified design is characterized by 'Differential pressure control' instead of temperature layer control'. This means a working principle that is completely independent of operational temperatures. The TVX-tank is replaced by a 'short cut' with a spring loaded checkvalve. One of the control valves creates a differential pressure that is too low to open the check valve but high enough to keep the next checkvalve in the next short cut closed. Measurements and computer simulations show that the pressure separator in the modified design works very well in district cooling applications. The temperature losses of heat

Emergency reactor container cooling facility

International Nuclear Information System (INIS)

Suzuki, Hiroaki; Matsumoto, Tomoyuki.

1992-01-01

The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

Emergency cooling system with hot-water jet pumps for nuclear reactors

International Nuclear Information System (INIS)

Reinsch, A.O.W.

1977-01-01

The ECCS for a PWR or BWR uses hot-water jet pumps to remove the thermal energy generated in the reactor vessel and stored in the water. The hot water expands in the nozzle part (Laval nozzle) of the jet pump and sucks in coolant (borated water) coming from a storage tank containing subcooled water. This water is mixing with the hot water/steam mixture from the Laval nozzle. The steam is condensed. The kinetic energy of the water is converted into a pressure increase which is sufficient to feed the water into the reactor vessel. The emergency cooling may further be helped by a jet condenser also operating according to the principle of a jet pump and condensing the steam generated in the reactor vessel. (DG) [de

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

Science.gov (United States)

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

2012-06-01

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

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

International Nuclear Information System (INIS)

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

1983-01-01

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

General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

International Nuclear Information System (INIS)

Roche, R.; Gaudez, J.C.

1964-01-01

In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

IMPROVEMENT OF SYSTEMS OF TECHNICAL WATER SUPPLY WITH COOLING TOWERS FOR STEAM POWER PLANTS TECHNICAL AND ECONOMIC INDICATORS PERFECTION. Part 1

Directory of Open Access Journals (Sweden)

Yu. A. Zenovich-Leshkevich-Olpinskiy

2016-01-01

Full Text Available In order to reduce the temperature of cooling water and increase the efficiency of use of power resources the main directions of modernization of systems of technical water supply with cooling towers at steam power plants are presented. The problems of operation of irrigation systems and water distribution systems of cooling towers are reviewed. The design of heat and mass transfer devices, their shortcomings and the impact on the cooling ability of the cooling tower are also under analysis. The use of droplet heat and mass transfer device based on the lattice polypropylene virtually eliminates the shortcomings of the film and droplet-film heat and mass transfer devices of the cooling tower, increasing lifetime, and improving the reliability and efficiency of the operation of the main equipment of thermal power plants. The design of the water distribution devices of cooling towers is also considered. It is noted that the most effective are water-spattering low-pressure nozzles made of polypropylene that provides uniform dispersion of water and are of a high reliability and durability.

Airfoil, platform, and cooling passage measurements on a rotating transonic high-pressure turbine

Science.gov (United States)

Nickol, Jeremy B.

An experiment was performed at The Ohio State University Gas Turbine Laboratory for a film-cooled high-pressure turbine stage operating at design-corrected conditions, with variable rotor and aft purge cooling flow rates. Several distinct experimental programs are combined into one experiment and their results are presented. Pressure and temperature measurements in the internal cooling passages that feed the airfoil film cooling are used as boundary conditions in a model that calculates cooling flow rates and blowing ratio out of each individual film cooling hole. The cooling holes on the suction side choke at even the lowest levels of film cooling, ejecting more than twice the coolant as the holes on the pressure side. However, the blowing ratios are very close due to the freestream massflux on the suction side also being almost twice as great. The highest local blowing ratios actually occur close to the airfoil stagnation point as a result of the low freestream massflux conditions. The choking of suction side cooling holes also results in the majority of any additional coolant added to the blade flowing out through the leading edge and pressure side rows. A second focus of this dissertation is the heat transfer on the rotor airfoil, which features uncooled blades and blades with three different shapes of film cooling hole: cylindrical, diffusing fan shape, and a new advanced shape. Shaped cooling holes have previously shown immense promise on simpler geometries, but experimental results for a rotating turbine have not previously been published in the open literature. Significant improvement from the uncooled case is observed for all shapes of cooling holes, but the improvement from the round to more advanced shapes is seen to be relatively minor. The reduction in relative effectiveness is likely due to the engine-representative secondary flow field interfering with the cooling flow mechanics in the freestream, and may also be caused by shocks and other

Numerical Simulation on Subcooled Boiling Heat Transfer Characteristics of Water-Cooled W/Cu Divertors

Science.gov (United States)

Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

2015-04-01

In order to realize safe and stable operation of a water-cooled W/Cu divertor under high heating condition, the exact knowledge of its subcooled boiling heat transfer characteristics under different design parameters is crucial. In this paper, subcooled boiling heat transfer in a water-cooled W/Cu divertor was numerically investigated based on computational fluid dynamic (CFD). The boiling heat transfer was simulated based on the Euler homogeneous phase model, and local differences of liquid physical properties were considered under one-sided high heating conditions. The calculated wall temperature was in good agreement with experimental results, with the maximum error of 5% only. On this basis, the void fraction distribution, flow field and heat transfer coefficient (HTC) distribution were obtained. The effects of heat flux, inlet velocity and inlet temperature on temperature distribution and pressure drop of a water-cooled W/Cu divertor were also investigated. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005), Funding of Jiangsu Innovation Program for Graduate Education (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe.

Science.gov (United States)

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes.

Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe

Science.gov (United States)

Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

2015-01-01

On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes. PMID:26201073

Thermal-hydraulic limitations on water-cooled limiters

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1984-08-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on current design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein concern primarily thermal protection, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

Studies on the behaviour of a passive containment cooling system for the Indian advanced heavy water reactor

International Nuclear Information System (INIS)

Maheshwari, N.K.; Saha, D.; Chandraker, D.K.; Kakodkar, A.; Venkat Raj, V.

2001-01-01

A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser's working is the potential degradation of immersed condenser's performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to understand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V 1 and V 2 reached almost steady value. With varying energy input rate V 1 pressure dropped below the pressure in V 2 . The system could efficiently purge air from V 1 to V 2 . The paper deals with the details of the tests conducted and the results obtained. (orig.) [de

Emergency core cooling system in BWR type reactors

International Nuclear Information System (INIS)

Takizawa, Yoji

1981-01-01

Purpose: To rapidly recover the water level in the reactor upon occurrence of slight leakages in the reactor coolant pressure boundary, by promoting the depressurization in the reactor to thereby rapidly increase the high pressure core spray flow rate. Constitution: Upon occurrence of reactor water level reduction, a reactor isolation cooling system and a high pressure core spray system are actuated to start the injection of coolants into a reactor pressure vessel. In this case, if the isolation cooling system is failed to decrease the flow rate in a return pipeway, flow rate indicators show a lower value as compared with a predetermined value. The control device detects it and further confirms the rotation of a high pressure spray pump to open a valve. By the above operation, coolants pumped by the high pressure spray pump is flown by way of a communication pipeway to the return pipeway and sprayed from the top of the pressure vessel. This allows the vapors on the water surface in the pressure vessel to be cooled rapidly and increases the depressurization effects. (Horiuchi, T.)

Comparison of enterotomy leak pressure among fresh, cooled, and frozen-thawed porcine jejunal segments.

Science.gov (United States)

Aeschlimann, Kimberly A; Mann, F A; Middleton, John R; Belter, Rebecca C

2018-05-01

OBJECTIVE To determine whether stored (cooled or frozen-thawed) jejunal segments can be used to obtain dependable leak pressure data after enterotomy closure. SAMPLE 36 jejunal segments from 3 juvenile pigs. PROCEDURES Jejunal segments were harvested from euthanized pigs and assigned to 1 of 3 treatment groups (n = 12 segments/group) as follows: fresh (used within 4 hours after collection), cooled (stored overnight at 5°C before use), and frozen-thawed (frozen at -12°C for 8 days and thawed at room temperature [23°C] for 1 hour before use). Jejunal segments were suspended and 2-cm enterotomy incisions were made on the antimesenteric border. Enterotomies were closed with a simple continuous suture pattern. Lactated Ringer solution was infused into each segment until failure at the suture line was detected. Leak pressure was measured by use of a digital transducer. RESULTS Mean ± SD leak pressure for fresh, cooled, and frozen-thawed segments was 68.3 ± 23.7 mm Hg, 55.3 ± 28.1 mm Hg, and 14.4 ± 14.8 mm Hg, respectively. Overall, there were no significant differences in mean leak pressure among pigs, but a significant difference in mean leak pressure was detected among treatment groups. Mean leak pressure was significantly lower for frozen-thawed segments than for fresh or cooled segments, but mean leak pressure did not differ significantly between fresh and cooled segments. CONCLUSIONS AND CLINICAL RELEVANCE Fresh porcine jejunal segments or segments cooled overnight may be used for determining intestinal leak pressure, but frozen-thawed segments should not be used.

Natural-circulation-cooling characteristics during PWR accident simulations

International Nuclear Information System (INIS)

Adams, J.P.; McCreery, G.E.; Berta, V.T.

1983-01-01

A description of natural circulation cooling characteristics is presented. Data were obtained from several pressurized water reactor accident simulations in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). The reliability of natural circulation cooling, its cooling effectiveness, and the effect of changing system conditions are described. Quantitative comparison of flow rates and time constants with theory for both single- and two-phase fluid conditions were made. It is concluded that natural circulation cooling can be relied on in plant recovery procedures in the absence of forced convection whenever the steam generator heat sink is available

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.

ANALISIS PENGGUNAAN WATER COOLED CONDENSER PADA MESIN PENGKONDISIAN UDARA PAKET (AC WINDOW

Directory of Open Access Journals (Sweden)

IKG Wirawan

2012-11-01

Full Text Available One of the important aspects in thermal design is refrigeration and air conditioning. Working principle of air conditioning is absorption and thermal dissipation process. Condenser is main component to release the heat from refrigerant to the cooling medium. In the present research, water cooled condenser was used to replace the commonly air condenser. Pressure and temperature at some section of the components were observed in order to examine the performance of the air conditioning system. The results showed that the COP varied from 9.66 to 12.4; refrigerationg effect varied from 1.31 kW to 1.86 kW; cooling capacity varied from 0.38 TR to 0.53 TR; and heat transfer varied from 2.2 kW to 2.98 kW.

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)

The characteristic of evaporative cooling magnet for ECRIS

Energy Technology Data Exchange (ETDEWEB)

Xiong, B., E-mail: xiongbin@mail.iee.ac.cn [Institute of Electrical Engineering, CAS, Beijing 100190 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Ruan, L.; Gu, G. B. [Institute of Electrical Engineering, CAS, Beijing 100190 (China); Lu, W.; Zhang, X. Z.; Zhan, W. L. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 73000 (China)

2016-02-15

Compared with traditional de-ionized pressurized-water cooled magnet of ECRIS, evaporative cooling magnet has some special characteristics, such as high cooling efficiency, simple maintenance, and operation. The analysis is carried out according to the design and operation of LECR4 (Lanzhou Electron Cyclotron Resonance ion source No. 4, since July 2013), whose magnet is cooled by evaporative cooling technology. The insulation coolant replaces the de-ionized pressurized-water to absorb the heat of coils, and the physical and chemical properties of coolant remain stable for a long time with no need for purification or filtration. The coils of magnet are immersed in the liquid coolant. For the higher cooling efficiency of coolant, the current density of coils can be greatly improved. The heat transfer process executes under atmospheric pressure, and the temperature of coils is lower than 70 °C when the current density of coils is 12 A/mm{sup 2}. On the other hand, the heat transfer temperature of coolant is about 50 °C, and the heat can be transferred to fresh air which can save cost of water cooling system. Two years of LECR4 stable operation show that evaporative cooling technology can be used on magnet of ECRIS, and the application advantages are very obvious.

The characteristic of evaporative cooling magnet for ECRIS

Science.gov (United States)

Xiong, B.; Ruan, L.; Gu, G. B.; Lu, W.; Zhang, X. Z.; Zhan, W. L.

2016-02-01

Compared with traditional de-ionized pressurized-water cooled magnet of ECRIS, evaporative cooling magnet has some special characteristics, such as high cooling efficiency, simple maintenance, and operation. The analysis is carried out according to the design and operation of LECR4 (Lanzhou Electron Cyclotron Resonance ion source No. 4, since July 2013), whose magnet is cooled by evaporative cooling technology. The insulation coolant replaces the de-ionized pressurized-water to absorb the heat of coils, and the physical and chemical properties of coolant remain stable for a long time with no need for purification or filtration. The coils of magnet are immersed in the liquid coolant. For the higher cooling efficiency of coolant, the current density of coils can be greatly improved. The heat transfer process executes under atmospheric pressure, and the temperature of coils is lower than 70 °C when the current density of coils is 12 A/mm2. On the other hand, the heat transfer temperature of coolant is about 50 °C, and the heat can be transferred to fresh air which can save cost of water cooling system. Two years of LECR4 stable operation show that evaporative cooling technology can be used on magnet of ECRIS, and the application advantages are very obvious.

Modification and application of water film model in COCOSYS for PWR's passive containment cooling

International Nuclear Information System (INIS)

Huang, Xi; Cheng, Xu

2014-01-01

Highlights: • Water film model in COCOSYS has been modified by considering film breakup. • Shear stress on film surface created by countercurrent flow has been considered. • Formation and development of rivulets have been taken into account. • Modified model has been applied for passive containment cooling system. • The modified water film model has optimized the simulation results. - Abstract: In this paper the physical model describing water film behaviors in German containment code system COCOSYS has been modified by taking into consideration the film breakup and subsequent phenomena as well as the effect of film interfacial shear stress created by countercurrent air flow. The modified model has extended its capability to predict particular water film behaviors including breakup at a critical film thickness based on minimum total energy criterion, the formation of rivulets according to total energy equilibrium as well as subsequent performance of rivulets according to several assumptions and observations from experiments. Furthermore, the modification considers also the change of velocity distribution on the cross section of film/rivulets due to shear stress. Based on the geometry of AP1000 and Generic Containment, simulations predicting containment pressure variation during accidents with operation of passive containment cooling system have been carried out. With the new model, considerably larger peak pressures are observed by comparing with those predicted with original water film model within a certain range of water film flow rate. Sensitivity analyses also point out that contact angle between water rivulets and steel substrate plays a significant role in the film cooling

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

A dual pressurized water reactor producing 2000 MWe

International Nuclear Information System (INIS)

Kang, K. M.; Suh, K. Y.

2010-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

Emergency core cooling system

International Nuclear Information System (INIS)

Abe, Nobuaki.

1993-01-01

A reactor comprises a static emergency reactor core cooling system having an automatic depressurization system and a gravitationally dropping type water injection system and a container cooling system by an isolation condenser. A depressurization pipeline of the automatic depressurization system connected to a reactor pressure vessel branches in the midway. The branched depressurizing pipelines are extended into an upper dry well and a lower dry well, in which depressurization valves are disposed at the top end portions of the pipelines respectively. If loss-of-coolant accidents should occur, the depressurization valve of the automatic depressurization system is actuated by lowering of water level in the pressure vessel. This causes nitrogen gases in the upper and the lower dry wells to transfer together with discharged steams effectively to a suppression pool passing through a bent tube. Accordingly, the gravitationally dropping type water injection system can be actuated faster. Further, subsequent cooling for the reactor vessel can be ensured sufficiently by the isolation condenser. (I.N.)

Water-ingress analysis for the 200 MWe pebble-bed modular high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Zheng Yanhua; Shi Lei; Wang Yan

2010-01-01

Water ingress into the primary circuit is generally recognized as one of the severe accidents with potential hazard to the modular high temperature gas-cooled reactor adopting steam-turbine cycle, which will cause a positive reactivity introduction, as well as the chemical corrosion of graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes and flammable water gas. The analysis of such a kind of important and particular accident is significant to verify the inherent safety characteristics of the modular HTR plants. Based on the preliminary design of the 200 MWe high temperature gas-cooled reactor pebble-bed modular (HTR-PM), the design basis accident of a double-ended guillotine break of one heating tube and the beyond design basis accident of a large break of the main steam collection plate have been analyzed by using TINTE code, which is a special transient analysis program for high temperature gas-cooled reactors. Some safety relevant concerns, such as the fuel temperature, the primary loop pressure, the graphite corrosion, the water gas releasing amount, as well as the natural convection influence on the condition of failing to close the blower flaps, have been studied in detail. The calculation results indicate that even under some severe hypothetical postulates, the HTR-PM is able to keep the inherent safeties of the modular high temperature gas-cooled reactor and has a relatively good natural plant response, which will not result in environmental radiation hazard.

Calculation of cooling tower plumes for high pressure wintry situations

International Nuclear Information System (INIS)

Gassmann, F.; Tinguely, M.; Haschke, D.

1982-12-01

The diffusion of the plumes of the projected nuclear power plants at Kaiseraugst and Schwoerstadt, during high pressure wintry conditions, has been examined using a mathematical model to simulate the plumes. For these calculations, microaerological measurements were made in the proximity of Kaiseraugst and Schwoerstadt. These give a typical image of the weather during high pressure wintry conditions, which is normally associated with an inversion, sometimes strong, at a low height. Dry cooling towers with natural draught, which offer an alternative solution to the wet cooling towers proposed for Kasieraugst, are examined equally. (Auth./G.T.H.)

Water-cooled U-tube grids for continuously operated neutral-beam injectors

International Nuclear Information System (INIS)

Hoffman, M.A.; Duffy, T.J.

1979-01-01

A design for water-cooled extractor grids for long-pulse and continuously operated ion sources for neutral-beam injectors is described. The most serious design problem encountered is that of minimizing the thermal deformation (bowing) of these slender grid rails, which have typical overall spans of 150 mm and diameters on the order of 1 mm. A unique U-tube design is proposed that offers the possibility of keeping the thermal bowing down to about 0.05 mm (about 2.0 mils). However, the design requires high-velocity cooling water at a Reynolds number of about 3 x 10 4 and an inlet pressure on the order of 4.67 x 10 6 Pa (677 psia) in order to keep the axial and circumferential temperature differences small enough to achieve the desired small thermal bowing. It appears possible to fabricate and assemble these U-tube grids out of molybdenum with high precision and with a reasonably small number of brazes

Cooling methods of station blackout scenario for LWR plants

International Nuclear Information System (INIS)

2012-01-01

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

IAEA activities in technology development for advanced water-cooled nuclear power plants

International Nuclear Information System (INIS)

Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

2003-01-01

As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

Dual pressurized light water reactor producing 2000 M We

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

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

Cardiovascular response to apneic immersion in cool and warm water

Science.gov (United States)

Folinsbee, L.

1974-01-01

The influence of prior exposure to cool water and the influence of lung volume on the responses to breath holding were examined. The bradycardia and vasoconstriction that occur during breath-hold diving in man are apparently the resultant of stimuli from apnea, relative expansion of the thorax, lung volume, esophageal pressure, face immersion, and thermal receptor stimulation. It is concluded that the bradycardia and vasoconstriction associated with breath holding during body immersion are not attenuated by a preexisting bradycardia and vasoconstriction due to cold.

Water cooled type nuclear power plant

International Nuclear Information System (INIS)

Arai, Shigeki.

1981-01-01

Purpose: To construct high efficiency a PWR type nuclear power plant with a simple structure by preparing high temperature and pressure water by a PWR type nuclear reactor and a pressurizer, converting the high temperature and high pressure water into steam with a pressure reducing valve and introducing the steam into a turbine, thereby generating electricity. Constitution: A pressurizer is connected downstream of a PWR type nuclear reactor, thereby maintaining the reactor at high pressure. A pressure-reducing valve is provided downstream of the pressurizer, the high temperature and pressure water is reduced in pressure, thereby producing steam. The steam is fed to a turbine, and electric power is generated by a generator connected to the turbine. The steam exhausted from the turbine is condensed by a condenser into water, and the water is returned through a feedwater heater to the reactor. Since the high temperature and pressure water in thus reduced in pressure thereby evaporating it, the steam can be more efficiently produced than by a steam generator. (Sekiya, K.)

Spacesuit Water Membrane Evaporator; An Enhanced Evaporative Cooling System for the Advanced Extravehicular Mobility Unit Portable Life Support System

Science.gov (United States)

Bue, Grant C.; Makinen, Janice V.; Miller, Sean; Campbell, Colin; Lynch, Bill; Vogel, Matt; Craft, Jesse; Wilkes, Robert; Kuehnel, Eric

2014-01-01

Development of the Advanced Extravehicular Mobility Unit (AEMU) portable life support subsystem (PLSS) is currently under way at NASA Johnson Space Center. The AEMU PLSS features a new evaporative cooling system, the Generation 4 Spacesuit Water Membrane Evaporator (Gen4 SWME). The SWME offers several advantages when compared with prior crewmember cooling technologies, including the ability to reject heat at increased atmospheric pressures, reduced loop infrastructure, and higher tolerance to fouling. Like its predecessors, Gen4 SWME provides nominal crew member and electronics cooling by flowing water through porous hollow fibers. Water vapor escapes through the hollow fiber pores, thereby cooling the liquid water that remains inside of the fibers. This cooled water is then recirculated to remove heat from the crew member and PLSS electronics. Test results from the backup cooling system which is based on a similar design and the subject of a companion paper, suggested that further volume reductions could be achieved through fiber density optimization. Testing was performed with four fiber bundle configurations ranging from 35,850 fibers to 41,180 fibers. The optimal configuration reduced the Gen4 SWME envelope volume by 15% from that of Gen3 while dramatically increasing the performance margin of the system. A rectangular block design was chosen over the Gen3 cylindrical design, for packaging configurations within the AEMU PLSS envelope. Several important innovations were made in the redesign of the backpressure valve which is used to control evaporation. A twin-port pivot concept was selected from among three low profile valve designs for superior robustness, control and packaging. The backpressure valve motor, the thermal control valve, delta pressure sensors and temperature sensors were incorporated into the manifold endcaps, also for packaging considerations. Flight-like materials including a titanium housing were used for all components. Performance testing

A design study of high electric power for fast reactor cooled by supercritical light water

International Nuclear Information System (INIS)

Koshizuka, Seiichi

2000-03-01

In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600 MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212 MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about 1.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power. (author)

Impact of the water symmetry factor on humidification and cooling strategies for PEM fuel cell stacks

Energy Technology Data Exchange (ETDEWEB)

Picot, D; Metkemeijer, R; Bezian, J J; Rouveyre, L [Centre d` Energetique, Ecole des Mines de Paris, 06 - Sophia Antipolis (France)

1998-10-01

In this paper, experimental water and thermal balances with three proton exchange membrane fuel cells (PEMFC) are proposed. On the test facility of Ecole des Mines de Paris, three De Nora SPA fuel cell stacks have been successfully studied: An 1 kW{sub e} prototype using Nafion {sup trademark} 117, a 5 and 10 kW{sub e} module using Nafion {sup trademark} 115. The averaged water symmetry factor determines strategies to avoid drying membrane. So, we propose analytical solutions to find compromises between humidification and cooling conditions, which determines outlet temperatures of gases. For transport applications, the space occupied by the power module must be reduced. One of the main efforts consists in decreasing the operative pressure. Thus, if adequate cooling power is applied, we show experimentally and theoretically the possibility to use De Nora PEM fuel cells with low pressure, without specific external humidification. (orig.)

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

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.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

International Nuclear Information System (INIS)

Yamada, K.; Aksan, S. N.

2012-01-01

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

Tests of cooling water pumps at Dukovany nuclear power plant

International Nuclear Information System (INIS)

Travnicek, J.

1986-01-01

Tests were performed to examine the operating conditions of the 1600 BQDV cooling pumps of the main coolant circuit of unit 1 of the Dukovany nuclear power plant. For the pumps, the performance was tested in the permissible operating range, points were measured below this range and the guaranteed operating point was verified. Pump efficiency was calculated from the measured values. The discussion of the measurement of parameters has not yet been finished because the obtained values of the amount delivered and thus of the pump efficiency were not up to expectation in all detail. It was also found that for obtaining the guaranteed flow the pump impeller had to be opened to 5deg -5.5deg instead of the declared 3deg. Also tested were pump transients, including the start of the pump, its stop, the operation and failure of one of the two pumps. In these tests, pressures were also measured at the inlet and the outlet of the inner part of the TG 11 turbine condenser. It was shown that the time course and the pressure course of the processes were acceptable. In addition to these tests, pressure losses in the condenser and the cooling water flow through the feed pump electromotor cooler wre tested for the case of a failure of one of the two pumps. (E.S.)

Investigation of the falling water flow with evaporation for the passive containment cooling system and its scaling-down criteria

Science.gov (United States)

Li, Cheng; Li, Junming; Li, Le

2018-02-01

Falling water evaporation cooling could efficiently suppress the containment operation pressure during the nuclear accident, by continually removing the core decay heat to the atmospheric environment. In order to identify the process of large-scale falling water evaporation cooling, the water flow characteristics of falling film, film rupture and falling rivulet were deduced, on the basis of previous correlation studies. The influences of the contact angle, water temperature and water flow rates on water converge along the flow direction were then numerically obtained and results were compared with the data for AP1000 and CAP1400 nuclear power plants. By comparisons, it is concluded that the water coverage fraction of falling water could be enhanced by either reducing the surface contact angle or increasing the water temperature. The falling water flow with evaporation for AP1000 containment was then calculated and the feature of its water coverage fraction was analyzed. Finally, based on the phenomena identification of falling water flow for AP1000 containment evaporation cooling, the scaling-down is performed and the dimensionless criteria were obtained.

European supercritical water cooled reactor (HPLWR Phase 2 project)

International Nuclear Information System (INIS)

Schulenberg, Thomas; Starflinger, Joerg; Marsault, Philippe; Bittermann, Dietmar; Maraczy, Czaba; Laurien, Eckart; Lycklama, Jan Aiso; Anglart, Henryk; Andreani, Michele; Ruzickova, Mariana; Heikinheimo, Liisa

2010-01-01

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 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners 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, housed fuel assemblies 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 innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. 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. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

Fabrication of gas turbine water-cooled composite nozzle and bucket hardware employing plasma spray process

Science.gov (United States)

Schilke, Peter W.; Muth, Myron C.; Schilling, William F.; Rairden, III, John R.

1983-01-01

In the method for fabrication of water-cooled composite nozzle and bucket hardware for high temperature gas turbines, a high thermal conductivity copper alloy is applied, employing a high velocity/low pressure (HV/LP) plasma arc spraying process, to an assembly comprising a structural framework of copper alloy or a nickel-based super alloy, or combination of the two, and overlying cooling tubes. The copper alloy is plamsa sprayed to a coating thickness sufficient to completely cover the cooling tubes, and to allow for machining back of the copper alloy to create a smooth surface having a thickness of from 0.010 inch (0.254 mm) to 0.150 inch (3.18 mm) or more. The layer of copper applied by the plasma spraying has no continuous porosity, and advantageously may readily be employed to sustain a pressure differential during hot isostatic pressing (HIP) bonding of the overall structure to enhance bonding by solid state diffusion between the component parts of the structure.

Dry cooling tower operating experience in the LOFT reactor

International Nuclear Information System (INIS)

Hunter, J.A.

1980-01-01

A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

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

GOTHIC code evaluation of alternative passive containment cooling features

International Nuclear Information System (INIS)

Gavrilas, M.; Todreas, E.N.; Driscoll, M.J.

1996-01-01

Reliance on passive cooling has become an important objective in containment design. Several reactor concepts have been set forth, which are equipped with entirely passively cooled containments. However, the problems that have to be overcome in rejecting the entire heat generated by a severe accident in a high-rating reactor (i.e. one with a rating greater than 1200 MW e ) have been found to be substantial and without obvious solutions. The GOTHIC code was verified and modified for containment cooling applications; optimal mesh sizes, computational time steps and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. The GOTHIC code was then employed to assess the effectiveness of several original heat rejection features that make it possible to cool high-rating containments. Two containment concepts were evaluated: one for a 1200 MW e new pressure tube light-water reactor, and one for a 1300 MW e pressurized-water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features has been predicted. The best-performance configurations-worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light-water reactor, and less than 0.45 MPa for the 1300 MW e pressurized-water reactor. (orig.)

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.

Exploration of Impinging Water Spray Heat Transfer at System Pressures Near the Triple Point

Science.gov (United States)

Golliher, Eric L.; Yao, Shi-Chune

2013-01-01

The heat transfer of a water spray impinging upon a surface in a very low pressure environment is of interest to cooling of space vehicles during launch and re-entry, and to industrial processes where flash evaporation occurs. At very low pressure, the process occurs near the triple point of water, and there exists a transient multiphase transport problem of ice, water and water vapor. At the impingement location, there are three heat transfer mechanisms: evaporation, freezing and sublimation. A preliminary heat transfer model was developed to explore the interaction of these mechanisms at the surface and within the spray.

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

Analysis of power and cooling cogeneration using ammonia-water mixture

International Nuclear Information System (INIS)

Padilla, Ricardo Vasquez; Demirkaya, Goekmen; Goswami, D. Yogi; Stefanakos, Elias; Rahman, Muhammad M.

2010-01-01

Development of innovative thermodynamic cycles is important for the efficient utilization of low-temperature heat sources such as solar, geothermal and waste heat sources. This paper presents a parametric analysis of a combined power/cooling cycle, which combines the Rankine and absorption refrigeration cycles, uses ammonia-water mixture as the working fluid and produces power and cooling simultaneously. This cycle, also known as the Goswami Cycle, can be used as a bottoming cycle using waste heat from a conventional power cycle or as an independent cycle using solar or geothermal energy. A thermodynamic study of power and cooling cogeneration is presented. The performance of the cycle for a range of boiler pressures, ammonia concentrations and isentropic turbine efficiencies are studied to find out the sensitivities of net work, amount of cooling and effective efficiencies. The roles of rectifier and superheater on the cycle performance are investigated. The cycle heat source temperature is varied between 90-170 o C and the maximum effective first law and exergy efficiencies for an absorber temperature of 30 o C are calculated as 20% and 72%, respectively. The turbine exit quality of the cycle for different boiler exit scenarios shows that turbine exit quality decreases when the absorber temperature decreases.

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.

Passive device for emergency core cooling of pressurized water reactors. Pasivno ustrojstvo za bezopasnost na vodo-voden atomen reaktor

Energy Technology Data Exchange (ETDEWEB)

Sikora, D

1984-02-28

The device proposed ensures additional margin of reactor subcriticality in case of post-accident emergency core cooling (ECC), using concentrated solution of chemical absorber and hot water from the secondary circuit. It consists of: a) a differential cylinder with a differential piston in it, with a lid and a seal, connected to a pipeline for secondary coolant; b) a pipeline for the secondary coolant; c) a volume between the lid and the piston for the secondary coolant from the steam generator; d) a discharge pipeline with a check valve of seal type connecting the inner volume of the differential cylinder to the discharge line; and e) a pipeline from the high-pressure volume of the differential cylinder filled with concentrated chemical absorber solution, to one of the main circulation loops. The device permits ECC innovation of the operating non-standard nuclear power plants with PWR type reactors.

Control rod driving hydraulic pressure device

International Nuclear Information System (INIS)

Ishida, Kazuo.

1990-01-01

Discharged water after actuating control rod drives in a BWR type reactor is once discharged to a discharging header, then returned to a master control unit and, subsequently, discharged to a reactor by way of a cooling water header. The radioactive level in the discharging header and the master control unit is increased by the reactor water to increase the operator's exposure. In view of the above, a riser is disposed for connecting a hydraulic pressure control unit incorporating a directional control valve and the cooling water head. When a certain control rod is inserted, the pressurized driving water is supplied through a hydraulic pressure control unit to the control rod drives. The discharged water from the control rod drives is entered by way of the hydraulic pressure control unit into the cooling water header and then returned to the reactor by way of other hydraulic pressure control unit and the control rod drives. Thus, the reactor water is no more recycled to the master control unit to reduce the radioactive exposure. (N.H.)

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.)

Head spray nozzle in reactor pressure vessel

International Nuclear Information System (INIS)

Hatano, Shun-ichi.

1990-01-01

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

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

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

Initial pressure spike and its propagation phenomena in sodium-water reaction tests for MONJU steam generators

International Nuclear Information System (INIS)

Sato, M.; Hiroi, H.; Tanaka, N.; Hori, M.

1977-01-01

With the objective of demonstrating the safe design of steam generators for prototype LMFBR MONJU against the postulated large-leak accident, a number of large-leak sodium-water reaction tests have been conducted using the SWAT-1 and SWAT-3 rigs. Investigation of the potential effects of pressure load on the system is one of the major concerns in these tests. This paper reports the behavior of initial pressure spike in the reaction vessel, its propagation phenomena to the simulated secondary cooling system, and the comparisons with the computer code for one-dimensional pressure wave propagation problems. Both rigs used are the scaled-down models of the helically coiled steam generators of MONJU. The SWAT-1 rig is a simplified model and consists of a reaction vessel (1/8 scale of MONJU evaporator with 0.4 m dia. and 2.5 m height) and a pressure relief system i.e., a pressure relief line and a reaction products tank. On the other hand, the SWAT-3 rig is a 1/2.5 scale of MONJU SG system and consists of an evaporator (reaction vessel with 1.3 m dia. and 6.35 m height), a superheater, an intermediate heat exchanger (IHX), a piping system simulating the secondary cooling circuit and a pressure relief system. The both water injection systems consist of a water injection line with a rupture disk installed in front of injection hole and an electrically heated water tank. Choice of water injection rates in the scaled-down models is made based on the method of iso-velocity modeling. Test results indicated that the characteristics of the initial pressure spike are dominated by those of initial water injection which are controlled by the conditions of water heater and the size of water injection hole, etc

Implementation of new core cooling monitoring system for light water reactors - BCCM (Becker Core Cooling Monitor)

International Nuclear Information System (INIS)

Coville, Patrick; Eliasson, Bengt; Stromqvist, Erik; Ward, Olav; Fox, Georges; Ashjian, D. T.

1998-01-01

Core cooling monitors are key instruments to protect reactors from large accidents due to loss of coolant. Sensors presented here are based on resistance thermometry. Temperature dependent resistance is powered by relatively high and constant current. Value of this resistance depends on thermal exchange with coolant and when water is no more surrounding the sensors a large increase of temperature is immediately generated. The same instrument can be operated with low current and will measure the local temperature up to 1260 o C in case of loss of coolant accident. Sensors are manufactured with very few components and materials already qualified for long term exposure to boiling or pressurized water reactors environment. Prototypes have been evaluated in a test loop up to 160 bars and in the Barsebaeck-1 reactor. Industrial sensors are now in operation in reactor Oskarshamn 2. (author)

Superconducting cable cooling system by helium gas at two pressures

International Nuclear Information System (INIS)

Dean, J.W.

1977-01-01

Thermally contacting, oppositely streaming, cryogenic fluid streams in the same enclosure in a closed cycle changes the fluid from a cool high pressure helium gas to a cooler reduced pressure helium gas in an expander so as to be at different temperature ranges and pressures respectively in go and return legs that are in thermal contact with each other and in thermal contact with a longitudinally extending superconducting transmission line enclosed in the same cable enclosure that insulates the line from the ambient at a temperature T 1 . By first circulating the fluid from a refrigerator at one end of the line as a cool gas at a temperature range T 2 to T 3 in the go leg, then circulating the gas through an expander at the other end of the line where the gas becomes a cooler gas at a reduced pressure and at a reduced temperature T 4 and finally by circulating the cooler gas back again to the refrigerator in a return leg at a temperature range T 4 to T 5 , while in thermal contact with the gas in the go leg, and in the same enclosure therewith for compression into a higher pressure gas at T 2 in a closed cycle, where T 2 greater than T 3 and T 5 greater than T 4 , the fluid leaves the enclosure in the go leg as a gas at its coldest point in the go leg, and the temperature distribution is such that the line temperature decreases along its length from the refrigerator due to the cooling from the gas in the return leg

Water-cooled non-thermal gliding arc for adhesion improvement of glass-fibre-reinforced polyester

DEFF Research Database (Denmark)

Kusano, Yukihiro; Sørensen, Bent F.; Løgstrup Andersen, Tom

2013-01-01

A non-equilibrium quenched plasma is prepared using a gliding-arc discharge generated between diverging electrodes and extended by a gas flow. It can be operated at atmospheric pressure and applied to plasma surface treatment to improve adhesion properties of material surfaces. In this work, glass......-fibre-reinforced polyester plates were treated using an atmospheric pressure gliding-arc discharge with air flow to improve adhesion with a vinylester adhesive. The electrodes were water-cooled so as to operate the gliding arc continually. The treatment improved wettability and increased the density of oxygen...

ITER cooling systems

International Nuclear Information System (INIS)

Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

1992-06-01

The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

Pressurized water reactor systems

International Nuclear Information System (INIS)

Meyer, P.J.

1975-01-01

Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

Emergency cooling system for a liquid metal cooled reactor

International Nuclear Information System (INIS)

Murata, Ryoichi; Fujiwara, Toshikatsu.

1980-01-01

Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

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

Water cycles in closed ecological systems: effects of atmospheric pressure.

Science.gov (United States)

Rygalov, Vadim Y; Fowler, Philip A; Metz, Joannah M; Wheeler, Raymond M; Bucklin, Ray A

2002-01-01

In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from ~1 to 10 L m-2 d-1 (~1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

Water cycles in closed ecological systems: effects of atmospheric pressure

Science.gov (United States)

Rygalov, Vadim Y.; Fowler, Philip A.; Metz, Joannah M.; Wheeler, Raymond M.; Bucklin, Ray A.; Sager, J. C. (Principal Investigator)

2002-01-01

In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from 1 to 10 L m-2 d-1 (1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

Investigation of Cooling Water Injection into Supersonic Rocket Engine Exhaust

Science.gov (United States)

Jones, Hansen; Jeansonne, Christopher; Menon, Shyam

2017-11-01

Water spray cooling of the exhaust plume from a rocket undergoing static testing is critical in preventing thermal wear of the test stand structure, and suppressing the acoustic noise signature. A scaled test facility has been developed that utilizes non-intrusive diagnostic techniques including Focusing Color Schlieren (FCS) and Phase Doppler Particle Anemometry (PDPA) to examine the interaction of a pressure-fed water jet with a supersonic flow of compressed air. FCS is used to visually assess the interaction of the water jet with the strong density gradients in the supersonic air flow. PDPA is used in conjunction to gain statistical information regarding water droplet size and velocity as the jet is broken up. Measurement results, along with numerical simulations and jet penetration models are used to explain the observed phenomena. Following the cold flow testing campaign a scaled hybrid rocket engine will be constructed to continue tests in a combusting flow environment similar to that generated by the rocket engines tested at NASA facilities. LaSPACE.

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

IMPROVEMENT OF SYSTEMS OF TECHNICAL WATER SUPPLY WITH COOLING TOWERS FOR HEAT POWER PLANTS TECHNICAL AND ECONOMIC INDICATORS PERFECTION. Part 2

Directory of Open Access Journals (Sweden)

Yu. A. Zenovich-Leshkevich-Olpinskiy

2016-01-01

Full Text Available The method of calculation of economic efficiency that can be universal and is suitable for feasibility study of modernization of irrigation and water distribution system of cooling towers has been developed. The method takes into account the effect of lower pressure exhaust steam in the condenser by lowering the temperature of the cooling water outlet of a cooling tower that aims at improvement of technical and economic indicators of heat power plants. The practical results of the modernization of irrigation and water distribution system of a cooling tower are presented. As a result, the application of new irrigation and water distribution systems of cooling towers will make it possible to increase the cooling efficiency by more than 4 оС and, therefore, to obtain the fuel savings by improving the vacuum in the turbine condensers. In addition, the available capacity of CHP in the summer period is increased. The results of the work, the experience of modernization of irrigation and water distribution systems of the Gomel CHP-2 cooling towers system, as well as the and methods of calculating of its efficiency can be disseminated for upgrading similar facilities at the power plants of the Belarusian energy system. Some measures are prosed to improve recycling systems, cooling towers and their structures; such measures might significantly improve the reliability and efficiency of technical water supply systems of heat power plants.

Corrosion inhibition measures in primary cooling water system during refurbishment of Cirus, re-commissioning and subsequent operation

International Nuclear Information System (INIS)

Rai, K.K.; Ramesh, N.; Sharma, R.C.

2008-01-01

Cirus is a 40 MWth, heavy water moderated, demineralized light water cooled, natural uranium fuelled research reactor. Reactor was commissioned in year 1960 and operated satisfactorily till 1990. After that availability factor started decreasing mainly due to equipment outage exhibiting signs of ageing. Based upon systematic ageing studies and assessment of condition of systems, structures and components, a refurbishment plan including safety upgrades was drawn up. Reactor was shut down in October 1997 for execution of jobs. After completion of refurbishment jobs reactor was started back in October 2002 and power operation was achieved in 2003. Primary cooling water (PCW) system consists of re-circulating pumps, heat exchangers, expansion tank, piping, valves, emergency storage reservoir (Ball Tank) and other components. Normally the fission heat from fuel is removed by re-circulating coolant in closed loop and transferred to seawater via heat exchangers. In case of outage of pumps, shut down cooling is provided by flow of water from Ball Tank under gravity to the underground dump tanks. The dissolved oxygen is maintained below 2 ppm and pH is maintained neutral to minimize corrosion of fuel cladding (Aluminum). This paper highlights the experience gained during segmentation of primary cooling water pipelines for pressure testing, measures taken to corrosion inhibition of primary cooling water lines to permit execution of refurbishment jobs, inspections and actions taken to repair/replace the corroded PCW pipe line segments, observations regarding corrosion related failures, re-commissioning of the system after refurbishment, assessment for safe reactor operation and experience during power operation. (author)

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.

Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

International Nuclear Information System (INIS)

Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

2003-01-01

The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

Design of a supercritical water-cooled reactor. Pressure vessel and internals

Energy Technology Data Exchange (ETDEWEB)

Fischer, Kai

2008-08-15

The High Performance Light Water Reactor (HPLWR) is a light water reactor with supercritical steam conditions which has been investigated within the 5th Framework Program of the European Commission. Due to the supercritical pressure of 25 MPa, water, used as moderator and as coolant, flows as a single phase through the core and can be directly fed to the turbine. Using the technology of coal fired power plants with supercritical steam conditions, the heat-up in the core is done in several steps to achieve the targeted high steam outlet temperature of 500.C without exceeding available cladding material limits. Based on a first design of a fuel assembly cluster for a HPLWR with a single pass core, the surrounding internals and the reactor pressure vessel (RPV) are dimensioned for the first time, following the safety standards of the nuclear safety standards commission in Germany. Furthermore, this design is extended to the incorporation of core arrangements with two and three passes. The design of the internals and the RPV are verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Additionally, a passive safety component for the feedwater inlet of the RPV of the HPLWR is designed. Its purpose is the reduction of the mass flow rate in case of a LOCA for a feedwater line break until further steps are executed. Starting with a simple vortex diode, several steps are executed to enhance the performance of the diode and adapt it to this application. Then, this first design is further optimized using combined 1D and 3D flow analyses. Parametric studies determine the performance and characteristic for changing mass flow rates for this backflow limiter. (orig.)

Fundamental research on the cooling characteristic of passive containment cooling system

International Nuclear Information System (INIS)

Kawakubo, M.; Kikura, H.; Aritomi, M.; Inaba, N.; Yamauchi, T.

2004-01-01

The objective of this experimental study is to clarify the heat transfer characteristics of the Passive Containment Cooling System (PCCS) with vertical heat transfer tubes for investigating the influence of non-condensable gas on condensation. Furthermore, hence we obtained new experimental correlation formula to calculate the transients in system temperature and pressure using the simulation program of the PCCS. The research was carried out using a forced circulation experimental loop, which simulates atmosphere inside PCCS with vertical heat transfer tubes if a loss of coolant accident (LOCA) occurs. The experimental facility consists of cooling water supply systems, an orifice flowmeter, and a tank equipped with the heat transfer pipe inside. Cooling water at a constant temperature is injected to the test part of heat transfer pipe vertically installed in the tank by forced circulation. At that time, the temperature of the cooling water between inlet and outlet of the pipe was measured to calculate the overall heat transfer coefficient between the cooling water and atmosphere in the tank. Thus, the heat transfer coefficient between heat transfer surface and the atmosphere in the tank considering the influence of the non-condensable gas was clarified. An important finding of this study is that the amount of condensation in the steamy atmosphere including non-condensable gas depends on the cooling water Reynolds number, especially the concentration of non-condensable gas that has great influence on the amount of condensation. (authors)

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

Model validation using CFD-grade experimental database for NGNP Reactor Cavity Cooling Systems with water and air

Energy Technology Data Exchange (ETDEWEB)

Manera, Annalisa [Univ. of Michigan, Ann Arbor, MI (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States); Petrov, Victor [Univ. of Michigan, Ann Arbor, MI (United States); Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Tompkins, Casey [Univ. of Wisconsin, Madison, WI (United States); Nunez, Daniel [Univ. of Michigan, Ann Arbor, MI (United States)

2018-02-13

This project has been focused on the experimental and numerical investigations of the water-cooled and air-cooled Reactor Cavity Cooling System (RCCS) designs. At this aim, we have leveraged an existing experimental facility at the University of Wisconsin-Madison (UW), and we have designed and built a separate effect test facility at the University of Michigan. The experimental facility at UW has underwent several upgrades, including the installation of advanced instrumentation (i.e. wire-mesh sensors) built at the University of Michigan. These provides highresolution time-resolved measurements of the void-fraction distribution in the risers of the water-cooled RCCS facility. A phenomenological model has been developed to assess the water cooled RCCS system stability and determine the root cause behind the oscillatory behavior that occurs under normal two-phase operation. Testing under various perturbations to the water-cooled RCCS facility have resulted in changes in the stability of the integral system. In particular, the effects on stability of inlet orifices, water tank volume have and system pressure been investigated. MELCOR was used as a predictive tool when performing inlet orificing tests and was able to capture the Density Wave Oscillations (DWOs) that occurred upon reaching saturation in the risers. The experimental and numerical results have then been used to provide RCCS design recommendations. The experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility has been equipped with state-of-theart high-resolution instrumentation to achieve so-called CFD grade experiments, that can be used for the validation of Computational Fluid Dynanmics (CFD) models, both RANS (Reynold-Averaged) and LES (Large Eddy Simulations). The effect of risers penetration in the upper plenum has been investigated as well.

Behavior of pressure rise and condensation caused by water evaporation under vacuum at high temperature

International Nuclear Information System (INIS)

Takase, Kazuyuki; Kunugi, Tomoaki; Yamazaki, Seiichiro; Fujii, Sadao

1998-01-01

Pressure rise and condensation characteristics during the ingress-of-coolant event (ICE) in fusion reactors were investigated using the preliminary ICE apparatus with a vacuum vessel (VV), an additional tank (AT) and an isolation valve (IV). A surface of the AT was cooled by water at RT. The high temperature and pressure water was injected into the VV which was heated up to 250degC and pressure and temperature transients in the VV were measured. The pressure increased rapidly with an injection time of the water because of the water evaporation. After the IV was opened and the VV was connected with the AT, the pressure in the VV decreased suddenly. From a series of the experiments, it was confirmed that control factors on the pressure rise were the flushing evaporation and boiling heat transfer in the VV, and then, condensation of the vapor after was effective to the depressurization in the VV. (author)

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

Experimental study of film media used for evaporative pre-cooling of air

International Nuclear Information System (INIS)

He, Suoying; Guan, Zhiqiang; Gurgenci, Hal; Hooman, Kamel; Lu, Yuanshen; Alkhedhair, Abdullah M.

2014-01-01

Highlights: • Two film media were experimentally studied in a low-speed wind tunnel. • Correlations for heat transfer coefficient and pressure drop were developed. • Cellulose media provide higher cooling efficiency and pressure drop than PVC media. • Water entrainment of PVC media happens even at relatively low air velocities. - Abstract: An open-circuit low-speed wind tunnel was used to study the performance of evaporative cooling with cellulose and Polyvinyl Chloride (PVC) corrugated media. These two film media were selected as part of a broader investigation on pre-cooling the entering air of natural draft dry cooling towers. The heat and mass transfer and pressure drop across the two media with three thicknesses (i.e., 100, 200 and 300 mm) were experimentally studied in the wind tunnel. The test data were non-dimensionalized and curve fitted to yield a set of correlations. It was found that the pressure drop range of the cellulose media is 1.5–101.7 Pa while the pressure drops of the PVC media are much lower with the range of 0.9–49.2 Pa, depending on the medium thickness, air velocity and water flow rate. The cooling efficiencies of the cellulose media vary from 43% to 90% while the cooling efficiencies of the PVC media are 8% to 65%, depending on the medium thickness and air velocity. The water entrainment off the media was detected by water sensitive papers, and found that the cellulose media have negligible water entrainment under the studied conditions while care must be taken in the use of PVC media as water entrainment happens even at relatively low air velocities

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

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

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

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

International Nuclear Information System (INIS)

Inasaka, Fujio; Adachi, Masaki; Aya, Izuo

2004-01-01

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

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...

Cooling facility of nuclear power plant

International Nuclear Information System (INIS)

Arai, Kenji; Nagasaki, Hideo.

1992-01-01

In a cooling device of a nuclear power plant, an exhaust pipe for an incondensible gas is branched. One of the branched exhaust pipes is opened in a pressure suppression pool water in a suppression chamber containing pool water and the other is opened at a lower portion of a dry well incorporating a pressure vessel. In a state where the pressure in the dry well is higher than that in the suppression chamber, an off-gas is exhausted effectively by way of the exhaustion pipe in communication with the suppression chamber. In a state where there is no difference between the pressures and the opening end of the exhaustion pipe in communication with the suppression chamber is sealed with water, off-gas is exhausted by way of the exhaustion pipe in communication with the lower portion of the dry well. Then, since the incondensible gas in a heat transfer pipe is not accumulated, after-heat can be removed efficiently. Satisfactory cooling is maintained even after the coincidence of the pressures in the dry well with that in the suppression chamber, to decrease a pressure in a reactor container. (N.H.)

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

Pressure control of a proton beam-irradiated water target through an internal flow channel-induced thermosyphon.

Science.gov (United States)

Hong, Bong Hwan; Jung, In Su

2017-07-01

A water target was designed to enhance cooling efficiency using a thermosyphon, which is a system that uses natural convection to induce heat exchange. Two water targets were fabricated: a square target without any flow channel and a target with a flow channel design to induce a thermosyphon mechanism. These two targets had the same internal volume of 8 ml. First, visualization experiments were performed to observe the internal flow by natural convection. Subsequently, an experiment was conducted to compare the cooling performance of both water targets by measuring the temperature and pressure. A 30-MeV proton beam with a beam current of 20 μA was used to irradiate both targets. Consequently, the target with an internal flow channel had a lower mean temperature and a 50% pressure drop compared to the target without a flow channel during proton beam irradiation. Copyright © 2017 Elsevier Ltd. All rights reserved.

Enhancement of LNG plant propane cycle through waste heat powered absorption cooling

International Nuclear Information System (INIS)

Rodgers, P.; Mortazavi, A.; Eveloy, V.; Al-Hashimi, S.; Hwang, Y.; Radermacher, R.

2012-01-01

In liquefied natural gas (LNG) plants utilizing sea water for process cooling, both the efficiency and production capacity of the propane cycle decrease with increasing sea water temperature. To address this issue, several propane cycle enhancement approaches are investigated in this study, which require minimal modification of the existing plant configuration. These approaches rely on the use of gas turbine waste heat powered water/lithium bromide absorption cooling to either (i) subcool propane after the propane cycle condenser, or (ii) reduce propane cycle condensing pressure through pre-cooling of condenser cooling water. In the second approach, two alternative methods of pre-cooling condenser cooling water are considered, which consist of an open sea water loop, and a closed fresh water loop. In addition for all cases, three candidate absorption chiller configurations are evaluated, namely single-effect, double-effect, and cascaded double- and single-effect chillers. The thermodynamic performance of each propane cycle enhancement scheme, integrated in an actual LNG plant in the Persian Gulf, is evaluated using actual plant operating data. Subcooling propane after the propane cycle condenser is found to improve propane cycle total coefficient of performance (COP T ) and cooling capacity by 13% and 23%, respectively. The necessary cooling load could be provided by either a single-effect, double-effect or cascaded and single- and double-effect absorption refrigeration cycle recovering waste heat from a single gas turbine operated at full load. Reducing propane condensing pressure using a closed fresh water condenser cooling loop is found result in propane cycle COP T and cooling capacity enhancements of 63% and 22%, respectively, but would require substantially higher capital investment than for propane subcooling, due to higher cooling load and thus higher waste heat requirements. Considering the present trend of short process enhancement payback periods in the

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...

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.

Pressure Effects on Solid State Phase Transformation of Aluminium Bronze in Cooling Process

International Nuclear Information System (INIS)

Hai-Yan, Wang; Jian-Hua, Liu; Gui-Rong, Peng; Yan, Chen; Yu-Wen, Liu; Fei, Li; Wen-Kui, Wang

2009-01-01

Effects of high pressure (6 GPa) on the solid state phase transformation kinetic parameters of aluminum bronze during the cooling process are investigated, based on the measurement and calculation of its solid state phase transformation temperature, duration and activation energy and the observation of its microstructures. The results show that high pressure treatment can reduce the solid phase transformation temperature and activation energy in the cooling process and can shorten the phase transformation duration, which is favorable when forming fine-grained aluminum bronze

3-Dimensional numerical study of cooling performance of a heat sink with air-water flow through mini-channel

Science.gov (United States)

Majumder, Sambit; Majumder, Abhik; Bhaumik, Swapan

2016-07-01

The present microelectronics market demands devices with high power dissipation capabilities having enhanced cooling per unit area. The drive for miniaturizing the devices to even micro level dimensions is shooting up the applied heat flux on such devices, resulting in complexity in heat transfer and cooling management. In this paper, a method of CPU processor cooling is introduced where active and passive cooling techniques are incorporated simultaneously. A heat sink consisting of fins is designed, where water flows internally through the mini-channel fins and air flows externally. Three dimensional numerical simulations are performed for large set of Reynolds number in laminar region using finite volume method for both developing flows. The dimensions of mini-channel fins are varied for several aspect ratios such as 1, 1.33, 2 and 4. Constant temperature (T) boundary condition is applied at heat sink base. Channel fluid temperature, pressure drop are analyzed to obtain best cooling option in the present study. It has been observed that as the aspect ratio of the channel decreases Nusselt number decreases while pressure drop increases. However, Nusselt number increases with increase in Reynolds number.

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.

Simulation of the injection system of cooling water to low pressure (Lpci) for a boiling water reactor (BWR) based on RELAP

International Nuclear Information System (INIS)

Delgado C, R. A.; Lopez S, E.; Chavez M, C.

2012-10-01

The present article describes the modeling and simulation of the Injection System of Cooling Water to Low Pressure (Lpci) for the nuclear power plant of Laguna Verde. Is very important to be able to predict the behavior of the nuclear plant in the case of an emergency stop, and while nearer to the reality are the results of a simulation, better is the safety protocol that can be devised. In the Engineering Faculty of the UNAM at the present is had logical models of the safety systems, but due to the nature of the same, these simulations do not provide of the quantity of enough information to be able to reproduce with more accuracy the behavior of the Lpci in the case of a severe accident. For this reason, the RELAP code was used for the flows modeling, components and structures of heat transfers in relation to the system Lpci. The modeling of the components is carried out with base on technical information of the nuclear plant and the results will be corroborated with information in reference documents as the Rasp (the Reactor analysis support package) and the Fsar (Final safety analysis report) for the nuclear power plant of Laguna Verde. (Author)

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.)

The research of materials and water chemistry for supercritical water-cooled reactors in Research Centre Rez

International Nuclear Information System (INIS)

Zychova, Marketa; Fukac, Rostislav; Vsolak, Rudolf; Vojacek, Ales; Ruzickova, Mariana; Vonkova, Katerina

2012-09-01

Research Centre Rez (CVR) is R and D company based in the Czech Republic. It was established as the subsidiary of the Nuclear Research Institute Rez plc. One of the main activities of CVR is the research of materials and chemistry for the generation IV reactor systems - especially the supercritical water-cooled one. For these experiments is CVR equipped by a supercritical water loop (SCWL) and a supercritical water autoclave (SCWA) serving for research of material and Supercritical Water-cooled Reactor (SCWR) environment compatibility experiments. SCWL is a research facility designed to material, water chemistry, radiolysis and other testing in SCWR environment, SCWA serves for complementary and supporting experiments. SCWL consists of auxiliary circuits (ensuring the required parameters as temperature, pressure and chemical conditions in the irradiation channel, purification and measurements) and irradiation channel (where specimens are exposed to the SCWR environment). The design of the loop is based on many years of experience with loop design for various types of corrosion/water chemistry experiments. Designed conditions in the test area of SCWL are 600 deg. C and 25 MPa. SCWL was designed in 2008 within the High Performance Light Water Reactor Phase 2 project and built during 2008 and 2009. The trial operations were performed in 2010 and 2011 and were divided into three phases - the first phase to verify the functionality of auxiliary circuits of the loop, the second phase to verify the complete facility (auxiliary circuits and functional irradiation channel internals) and the third phase to verify the feasibility of corrosion tests with the complete equipment and specimens. All three trial operations were very successful - designed conditions and parameters were reached. (authors)

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.

Pressure tube type reactors

International Nuclear Information System (INIS)

Komada, Masaoki.

1981-01-01

Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

Thermal-hydraulic analysis on the whole module of water cooled ceramic breeder blanket for CFETR

Energy Technology Data Exchange (ETDEWEB)

Jiang, Kecheng; Ma, Xuebin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China); Cheng, Xiaoman [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Lin, Shuang [University of Science and Technology of China, Hefei, Anhui, 230027 (China); Huang, Kai [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); Liu, Songlin, E-mail: slliu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui, 230031 (China); University of Science and Technology of China, Hefei, Anhui, 230027 (China)

2016-11-15

Highlights: • The 3D thermal hydraulic analysis on the whole module of WCCB is performed by CFD method. • Temperature field and mass flow distribution have been obtained. • The design of WCCB is reasonable from the perspective of thermal-hydraulics. • The scheme for further optimization has been proposed. - Abstract: The Water Cooled Ceramic Breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). The thermal-hydraulic analysis is essential because the blanket should remove the high heat flux from the plasma and the volumetric heat generated by neutrons. In this paper, the detailed three dimensional (3D) thermal hydraulic analysis on the whole module of WCCB blanket has been performed by Computational Fluid Dynamics (CFD) method, which is capable of solving conjugate heat transfer between solid structure and fluid. The main results, including temperature field, distribution of mass flow rate and coolant pressure drop, have been calculated simultaneously. These provides beneficial guidance data for the further structural optimization and for the design arrangement of primary and secondary circuit. Under the total heat source of 1.23 MW, the coolant mass flow rate of 5.457 kg/s is required to make coolant water corresponding to the Pressurized Water Reactor (PWR) condition (15.5 MPa, 285 °C–325 °C), generating the total coolant pressure drop (△P) of 0.467 MPa. The results show that the present structural design can make all the materials effectively cooled to the allowable temperature range, except for a few small modifications on the both sides of FW. The main components, including the first wall (FW), cooling plates (CPs), side wall (SWs)&stiffening plates (SPs) and the manifold(1–4), dominate 4.7%/41.7%/13%/40.6% of the total pressure drop, respectively. Additionally, the mass flow rate of each channel has been obtained, showing the peak relative deviation of 3.4% and 2% from the average for the paratactic

Hydrogen co-production from subcritical water-cooled nuclear power plants in Canada

Energy Technology Data Exchange (ETDEWEB)

Gnanapragasam, N.; Ryland, D.; Suppiah, S., E-mail: gnanapragasamn@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

2013-06-15

Subcritical water-cooled nuclear reactors (Sub-WCR) operate in several countries including Canada providing electricity to the civilian population. The high-temperature-steam-electrolysis process (HTSEP) is a feasible and laboratory-demonstrated large-scale hydrogen-production process. The thermal and electrical integration of the HTSEP with Sub-WCR-based nuclear-power plants (NPPs) is compared for best integration point, HTSEP operating condition and hydrogen production rate based on thermal energy efficiency. Analysis on integrated thermal efficiency suggests that the Sub-WCR NPP is ideal for hydrogen co-production with a combined efficiency of 36%. HTSEP operation analysis suggests that higher product hydrogen pressure reduces hydrogen and integrated efficiencies. The best integration point for the HTSEP with Sub-WCR NPP is upstream of the high-pressure turbine. (author)

Solar heating, cooling, and domestic hot water system installed at Kaw Valley State Bank and Trust Company, Topeka, Kansas

Science.gov (United States)

1980-01-01

The building has approximately 5600 square feet of conditioned space. Solar energy was used for space heating, space cooling, and preheating domestic hot water (DHW). The solar energy system had an array of evacuated tube-type collectors with an area of 1068 square feet. A 50/50 solution of ethylene glycol and water was the transfer medium that delivered solar energy to a tube-in-shell heat exchanger that in turn delivered solar heated water to a 1100 gallon pressurized hot water storage tank. When solar energy was insufficient to satisfy the space heating and/or cooling demand, a natural gas-fired boiler provided auxiliary energy to the fan coil loops and/or the absorption chillers. Extracts from the site files, specification references, drawings, and installation, operation and maintenance instructions are presented.

CAREM 25: Suppression pool cooling and purification system

International Nuclear Information System (INIS)

Carlevaris, Rodolfo; Patrignani, Alberto; Vindrola, Carlos; Palmerio, Hector D.; Quiroz, Horacio; Ramilo, Lucia B.

2000-01-01

The suppression pool cooling and purification system has the following main functions: purify and cool water from the suppression pool, cool and send water to the residual heat extraction system, and transfer water to the fuel element transference channel. In case of Loss of Coolant Accident (LOCA), the system sends water from the suppression pool to the spray network, thus cooling and reducing pressure in the primary containment. The system has been designed in accordance with the requirements of the following standards: ANSI/ANS 52.1; ANSI/ANS 57.2; ANSI/ANS 56.2; ANSI/ANS 59.1; ANSI/ANS 58.3; ANSI/ANS 58.9; and ANSI/ANS 56.5. The design of the system fulfils all the assigned functions. (author)

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.

Innovative technologies for Faraday shield cooling

International Nuclear Information System (INIS)

Rosenfeld, J.H.; Lindemuth, J.E.; North, M.T.; Goulding, R.H.

1995-01-01

Alternative advanced technologies are being evaluated for use in cooling the Faraday shields used for protection of ion cyclotron range of frequencies (ICR) antennae in Tokamaks. Two approaches currently under evaluation include heat pipe cooling and gas cooling. A Monel/water heat pipe cooled Faraday shield has been successfully demonstrated. Heat pipe cooling offers the advantage of reducing the amount of water discharged into the Tokamak in the event of a tube weld failure. The device was recently tested on an antenna at Oak Ridge National Laboratory. The heat pipe design uses inclined water heat pipes with warm water condensers located outside of the plasma chamber. This approach can passively remove absorbed heat fluxes in excess of 200 W/cm 2 ;. Helium-cooled Faraday shields are also being evaluated. This approach offers the advantage of no liquid discharge into the Tokamak in the event of a tube failure. Innovative internal cooling structures based on porous metal cooling are being used to develop a helium-cooled Faraday shield structure. This approach can dissipate the high heat fluxes typical of Faraday shield applications while minimizing the required helium blower power. Preliminary analysis shows that nominal helium flow and pressure drop can sufficiently cool a Faraday shield in typical applications. Plans are in progress to fabricate and test prototype hardware based on this approach

Quantum limits of photothermal and radiation pressure cooling of a movable mirror

International Nuclear Information System (INIS)

Pinard, M; Dantan, A

2008-01-01

We present a general quantum-mechanical theory for the cooling of a movable mirror in an optical cavity when both radiation pressure self-cooling and photothermal cooling effects are present, and show that these two mechanisms may bring the oscillator close to its quantum ground state, although in quite different regimes. Self-cooling caused by coherent exchange of excitations between the cavity mode and the mirror vibrational mode is shown to dominate in the good-cavity regime-when the mechanical resonance frequency is larger than the cavity decay rate, whereas photothermal-induced cooling can be made predominant in the bad-cavity limit. Both situations are compared, and the relevant physical quantities to be optimized in order to reach the lowest final excitation number states are extracted.

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

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

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-...

GOTHIC code evaluation of alternative passive containment cooling features

International Nuclear Information System (INIS)

Gavrilas, M.; Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1994-01-01

The GOTHIC code was employed to assess the effectiveness of several original heat rejection features that make it possible to cool large rating containments. The code was first verified and modified for specific containment cooling applications; optimal mesh sizes, computational time steps, and applicable heat transfer correlations were examined. The effect of the break location on circulation patterns that develop inside the containment was also evaluated. GOTHIC was then used to obtain performance predictions for two containment concepts: a 1200 MW e new pressure tube light water reactor, and a 1300 MW e pressurized water reactor. The effectiveness of various containment configurations that include specific pressure-limiting features have been predicted. For the 1200 MW e pressure tube light water reactor, the evaluated pressure-limiting features are: a large water pool connected to the calandria, large containment free volume and an air-convection annulus. For the 1300 MW e pressurized water reactor, an external moat, an internal water pool, and an air-convection annulus were evaluated. The performance of the proposed containment configurations is dependent on the extent of thermal stratification inside the containment. The best-performance configurations/worst-case-accident scenarios that were examined yielded peak pressures of less than 0.30 MPa for the 1200 MW e pressure tube light water reactor, and less than 0.45 MPa for the 1300 MW e pressurized water reactor. The low peak pressure predicted for the 1200 MW e pressure tube light water reactor can be in part attributed to its relatively large free volume, while the relatively high peak pressure predicted for the 1300 MW e pressurized water reactor can be attributed to its relatively small free volume (i.e., the size used was that of a pressurized water reactor containment designed with active heat removal features). (author)

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.

Evaluation on the heat removal capacity of the first wall for water cooled breeder blanket of CFETR

Energy Technology Data Exchange (ETDEWEB)

Jiang, Kecheng, E-mail: jiangkecheng@ipp.ac.cn; Cheng, Xiaoman; Chen, Lei; Huang, Kai; Ma, Xuebin; Liu, Songlin

2016-02-15

Highlights: • Heat removal capacity of the FW is evaluated under BWR, PWR and He coolant inlet conditions. • Heat transfer property of the gas–liquid two phase and the two boiling crises are analyzed. • Heat removal capacity of water is larger than helium coolant. - Abstract: The water cooled ceramic breeder blanket (WCCB) is being researched for Chinese Fusion Engineering Test Reactor (CFETR). As an important component of the blanket, the FW should satisfy with the thermal requirements in any case. In this paper, three parameters including the heat removal capacity, coolant pressure drop as well as the temperature rise of the FW were investigated under different coolant velocity and heat flux from the plasma. Using the same first wall structure, two main water cooled schemes including Boiling Water Reactor (BWR, 7 MPa pressure and 265 °C temperature inlet) and Pressurized Water Reactor (PWR, 15 MPa pressure and 285 °C temperature inlet) conditions are discussed in the thermal hydraulic calculation. For further research, the thermal hydraulic characteristics of using helium as coolant (8 MPa pressure, 300 °C temperature inlet) are also explored to provide CFETR blanket design with more useful data supports. Without regard to the outlet coolant condition requirements of the blanket, the results indicate that the ultimate heat flux that the FW can resist is 2.2 MW/m{sup 2} at velocity of 5 m/s for BWR, 2.0 MW/m{sup 2} at velocity of 5 m/s for PWR and 0.87 MW/m{sup 2} for helium at velocity 100 m/s under the chosen operation condition. The detrimental departure from nucleate boiling (DNB) crisis would occur at the velocity of 1 m/s under the heat flux of 3 MW/m{sup 2} and dry out crisis appears at the velocity of less than 0.2 m/s with the heat flux of more than 1 MW/m{sup 2} for BWR. The further blanket/FW optimization design is provided with more useful data references according to the abundant calculation results.

Cooling system upon reactor isolation

International Nuclear Information System (INIS)

Yamamoto, Kohei; Oda, Shingo; Miura, Satoshi

1992-01-01

A water level indicator for detecting the upper limit value for a range of using a suppression pool and a thermometer for detecting the temperature of water at the cooling water inlet of an auxiliary device are disposed. When a detection signal is intaken and the water level in the suppression pool reach the upper limit value for the range of use, a secondary flow rate control value is opened and a primary flow rate control valve is closed. When the temperature of the water at the cooling water inlet of the auxiliary device reaches the upper limit value, the primary and the secondary flow rate control valves are opened. During a stand-by state, the first flow rate control valve is set open and the secondary flow rate control valve is set closed respectively. After reactor isolation, if a reactor water low level signal is received, an RCIC pump is actuated and cooling water is sent automatically under pressure from a condensate storage tank to the reactor and the auxiliary device requiring coolants by way of the primary flow rate control valve. Rated flow rate is ensured in the reactor and cooling water of an appropriate temperature can be supplied to the auxiliary device. (N.H.)

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

Emergency core cooling systems

International Nuclear Information System (INIS)

Kubokoya, Takashi; Okataku, Yasukuni.

1984-01-01

Purpose: To maintain the fuel soundness upon loss of primary coolant accidents in a pressure tube type nuclear reactor by injecting cooling heavy water at an early stage, to suppress the temperature of fuel cans at a lower level. Constitution: When a thermometer detects the temperature rise and a pressure gauge detects that the pressure for the primary coolants is reduced slightly from that in the normal operation upon loss of coolant accidents in the vicinity of the primary coolant circuit, heavy water is caused to flow in the heavy water feed pipeway by a controller. This enables to inject the heavy water into the reactor core in a short time upon loss of the primary coolant accidents to suppress the temperature rise in the fuel can thereby maintain the fuel soundness. (Moriyama, K.)

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1976-01-01

Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

DUSEL Facility Cooling Water Scaling Issues

Energy Technology Data Exchange (ETDEWEB)

Daily, W D

2011-04-05

Precipitation (crystal growth) in supersaturated solutions is governed by both kenetic and thermodynamic processes. This is an important and evolving field of research, especially for the petroleum industry. There are several types of precipitates including sulfate compounds (ie. barium sulfate) and calcium compounds (ie. calcium carbonate). The chemical makeup of the mine water has relatively large concentrations of sulfate as compared to calcium, so we may expect that sulfate type reactions. The kinetics of calcium sulfate dihydrate (CaSO4 {center_dot} 2H20, gypsum) scale formation on heat exchanger surfaces from aqueous solutions has been studied by a highly reproducible technique. It has been found that gypsum scale formation takes place directly on the surface of the heat exchanger without any bulk or spontaneous precipitation in the reaction cell. The kinetic data also indicate that the rate of scale formation is a function of surface area and the metallurgy of the heat exchanger. As we don't have detailed information about the heat exchanger, we can only infer that this will be an issue for us. Supersaturations of various compounds are affected differently by temperature, pressure and pH. Pressure has only a slight affect on the solubility, whereas temperature is a much more sensitive parameter (Figure 1). The affect of temperature is reversed for calcium carbonate and barium sulfate solubilities. As temperature increases, barium sulfate solubility concentrations increase and scaling decreases. For calcium carbonate, the scaling tendencies increase with increasing temperature. This is all relative, as the temperatures and pressures of the referenced experiments range from 122 to 356 F. Their pressures range from 200 to 4000 psi. Because the cooling water system isn't likely to see pressures above 200 psi, it's unclear if this pressure/scaling relationship will be significant or even apparent. The most common scale minerals found in the

Performance comparison between ethanol phase-change immersion and active water cooling for solar cells in high concentrating photovoltaic system

International Nuclear Information System (INIS)

Wang, Yiping; Wen, Chen; Huang, Qunwu; Kang, Xue; Chen, Miao; Wang, Huilin

2017-01-01

Highlights: • Thermal performances of ethanol phase-change immersion and active water cooling are compared. • Effects of operation parameters on ethanol phase-change immersion are studied. • Optimum filling ratio is 30% for ethanol phase-change immersion cooling system. • Exergy efficiency of ethanol phase-change immersion method increases by 57%. - Abstract: This paper presents an optimized ethanol phase-change immersion cooling method to obtain lower temperature of dense-array solar cells in high concentrating photovoltaic system. The thermal performances of this system were compared with a conventional active water cooling system with minichannels from the perspectives of start-up characteristic, temperature uniformity, thermal resistance and heat transfer coefficient. This paper also explored the influences of liquid filling ratio, absolute pressure and water flow rate on thermal performances. Dense-array LEDs were used to simulate heat power of solar cells worked under high concentration ratios. It can be observed that the optimal filling ratio was 30% in which the thermal resistance was 0.479 °C/W and the heat transfer coefficient was 9726.21 W/(m 2 ·°C). To quantify the quality of energy output of two cooling systems, exergy analysis are conducted and maximum exergy efficiencies were 17.70% and 11.27%, respectively. The experimental results represent an improvement towards thermal performances of ethanol phase-change immersion cooling system due to the reduction in contact thermal resistance. This study improves the operation control and applications for ethanol phase-change immersion cooling technology.

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.

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)

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.

Estimation on the Pressure Loss of the Conceptual Primary Cooling System and Design of the Primary Cooling Pump for a Research Reactor

International Nuclear Information System (INIS)

Seo, Kyoung Woo; Oh, Jae Min; Park, Jong Hark; Chae, Hee Taek; Seo, Jae Kwang; Park, Cheon Tae; Yoon, Ju Hyeon; Lee, Doo Jeong

2009-01-01

A new conceptual primary cooling system (PCS) for a research reactor has been designed for an adequate cooling to the reactor core which has various powers ranging from 30MW through 80MW. The developed primary cooling system consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. Because the system flow rate should be determined by the thermal hydraulic design analysis for the core, the heads to design the primary cooling pumps (PCPs) in a PCS will be estimated by the variable system flow rates. The heads of the part of a research reactor vessel was evaluated by the previous study. The various pressure losses of the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. The purpose of this research is to estimate the various pressure losses and to design the PCPs

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 ...

Absorption of water vapour in the falling film of water-(LiBr + LiI + LiNO{sub 3} + LiCl) in a vertical tube at air-cooling thermal conditions

Energy Technology Data Exchange (ETDEWEB)

Bourouis, Mahmoud; Valles, Manel; Medrano, Marc; Coronas, Alberto [Centro de Innovacion Tecnologica en Revalorizacion Energetica y Refrigeracion, CREVER, Universitat Rovira i Virgili, Autovia de Salou, s/n, 43006, Tarragona (Spain)

2005-05-01

In air-cooled water-LiBr absorption chillers the working conditions in the absorber and condenser are shifted to higher temperatures and concentrations, thereby increasing the risk of crystallisation. To develop this technology, two main problems are to be addressed: the availability of new salt mixtures with wider range of solubility than water-LiBr, and advanced absorber configurations that enable to carry out simultaneously an appropriate absorption process and an effective air-cooling. One way of improving the solubility of LiBr aqueous solutions is to add other salts to create multicomponent salt solutions. The aqueous solution of the quaternary salt system (LiBr + LiI + LiNO{sub 3} + LiCl) presents favourable properties required for air-cooled absorption systems: less corrosive and crystallisation temperature about 35 K lower than that of water-LiBr.This paper presents an experimental study on the absorption of water vapour over a wavy laminar falling film of an aqueous solution of (LiBr + LiI + LiNO{sub 3} + LiCl) on the inner wall of a water-cooled smooth vertical tube. Cooling water temperatures in the range 30-45 C were selected to simulate air-cooling thermal conditions. The results are compared with those obtained in the same experimental set-up with water-LiBr solutions.The control variables for the experimental study were: absorber pressure, solution Reynolds number, solution concentration and cooling water temperature. The parameters considered to assess the absorber performance were: absorber thermal load, mass absorption flux, degree of subcooling of the solution leaving the absorber, and the falling film heat transfer coefficient.The higher solubility of the multicomponent salt solution makes possible the operation of the absorber at higher salt concentration than with the conventional working fluid water-LiBr. The absorption fluxes achieved with water-(LiBr + LiI + LiNO{sub 3} + LiCl) at a concentration of 64.2 wt% are around 60 % higher than

Thermal-hydraulic limitations on water-cooled fusion reactor components

International Nuclear Information System (INIS)

Cha, Y.S.; Misra, B.

1986-01-01

An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

CAREM-25. Suppression Pool Cooling and Purification System

International Nuclear Information System (INIS)

Carlevaris, Rodolfo; Palmerio, D.; Patrignani, A.; Quiroz, H.; Ramilo, L.; Vindrola, C.

2000-01-01

The Suppression Pool Cooling and Purification System has the following main functions: purify and cool water from the Suppression Pool, cool and send water to the Residual Heat Extraction System, and transfer water to the Fuel Element Transference Channel. In case of Loss of Coolant Accident (LOCA), the system sends water from the Suppression Pool to the spray network, thus cooling and reducing pressure in the primary containment.The system has been designed in accordance with the requirements of the following standards ANSI/ANS 52.1 [1], ANSI/ANS 57.2 [2], ANSI/ANS 56.2 [3], ANSI/ANS 59.1 [4] ANSI/ANS 58.3 [5], ANSI/ANS 58.9 [6], and ANSI/ANS 56.5 [7]. The design of the system fulfils all the assigned functions

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

A feasibility study on feed and bleed for pressurized water reactors

International Nuclear Information System (INIS)

Yi-Shung Chen; Shimeck, D.J.; Sullivan, L.H.

1983-01-01

By injecting coolant with a high pressure emergency core cooling system, and removing the heated/ vaporized fluid by way of the pressurizer power operated relief valve, primary feed and bleed cooling denotes an operation whereby reactor core cooling is maintained. This paper presents the results from an experimental and analytical study that includes a simplified analysis of mass and energy balances associated with the feed and bleed, examination of test data from the Semiscale system, RELAP5 code analyses of both Semiscale and a four-loop Westinghouse plant, and the primary coolant system behavior for a transient that leads to the need for feed and bleed. Examination of the parameters that govern a stable feed and bleed operation identifies four key parameters such as: core decay heat, cooling water injection capacity, power operated relief valve (PORV) energy removal rate, and PORV mass removal rate. A simplified analytical approach to determining if stable feed and bleed is feasible, has been developed and corroborated by experimental data and computer code calculations. The Semiscale tests have not only provided test data for code assessment, but also have identified the factors influencing the PORV discharge, which is the most variable of the boundary conditions influencing feed and bleed. The RELAP5 computer code has demonstrated the capability of predicting the Semiscale experiments, and when applied to a four-loop Westinghouse plant has indicated that primary feed and bleed is a viable cooling mechanism. This has also been shown by using the simplified analytical method

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.

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Simultaneous heat and mass transfer to air from a compact heat exchanger with water spray precooling and surface deluge cooling

International Nuclear Information System (INIS)

Zhang, Feini; Bock, Jessica; Jacobi, Anthony M.; Wu, Hailing

2014-01-01

Various methods are available to enhance heat exchanger performance with evaporative cooling. In this study, evaporative mist precooling, deluge cooling, and combined cooling schemes are examined experimentally and compared to model predictions. A flexible model of a compact, finned-tube heat exchanger with a wetted surface is developed by applying the governing conservation and rate equations and invoking the heat and mass transfer analogy. The model is applicable for dry, partially wet, or fully wet surface conditions and capable of predicting local heat/mass transfer, wetness condition, and pressure drop of the heat exchanger. Experimental data are obtained from wind tunnel experiments using a louver-fin flat-tube heat exchanger with single-phase tube-side flow. Total capacity, pressure drop, and water drainage behavior under various water usage rates and air face velocities are analyzed and compared to data for dry-surface conditions. A heat exchanger partitioning method for evaporative cooling is introduced to study partially wet surface conditions, as part of a consistent and general method for interpreting wet-surface performance data. The heat exchanger is partitioned into dry and wet portions by introducing a wet surface factor. For the wet part, the enthalpy potential method is used to determine the air-side sensible heat transfer coefficient. Thermal and hydraulic performance is compared to empirical correlations. Total capacity predictions from the model agree with the experimental results with an average deviation of 12.6%. The model is also exercised for four water augmentation schemes; results support operating under a combined mist precooling and deluge cooling scheme. -- Highlights: • A new spray-cooled heat exchanger model is presented and is validated with data. • Heat duty is shown to be asymptotic with spray flow rate. • Meaningful heat transfer coefficients for partially wet conditions are obtained. • Colburn j wet is lower than j dry

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

Pressure balanced type membrane covered polarographic oxygen detectors for use in high temperature-high pressure water, (1)

International Nuclear Information System (INIS)

Nakayama, Norio; Uchida, Shunsuke

1984-01-01

A pressure balanced type membrane covered polarographic oxygen detector was developed to determine directly oxygen concentrations in high temperature, high pressure water without cooling and pressure reducing procedures. The detector is characterized by the following features: (1) The detector body and the membrane for oxygen penetration are made of heat resistant resin. (2) The whole detector body is contained in a pressure chamber where interior and exterior pressures of the detector are balanced. (3) Thermal expansion of the electrolyte is absorbed by deformation of a diaphragm attached to the detector bottom. (4) The effect of dissolved Ag + on the signal current is eliminated by applying a guard electrode. As a result of performance tests at elevated temperature, it was demonstrated that a linear relationship between oxygen concentration and signal current was obtained up to 285 0 C, which was stabilized by the guard electrode. The minimum O 2 concentration detectable was 0.03ppm (9.4 x 10 -7 mol/kg). (author)

Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

International Nuclear Information System (INIS)

Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence

2012-01-01

Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

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)

Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

International Nuclear Information System (INIS)

Schumann, S.

1984-01-01

The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.) [de

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)

Conceptual design of a high temperature water-cooled divertor for a fusion power reactor

International Nuclear Information System (INIS)

Giancarli, L.; Bonal, J.P.; Puma, A. Li; Michel, B.; Sardain, P.; Salavy, J.F.

2005-01-01

This paper presents the conceptual design of a water-cooled divertor target using EUROFER as structural material, water coolant pressure and outlet temperature, respectively, of 15.5 MPa and 325 o C, and W-alloy monoblocks as armour. Assuming an advanced interface, formed by a thermal barrier in the pipe front part and a compliance layer between W and steel, this concept is able to withstand an incident surface heat flux of 15 MW/m 2 . Both thermal barrier and compliance layer are made of carbon-based materials. The main issues are the manufacturing process of the steel/W interface, and the behaviour under irradiation of graphite materials

Conceptual design of a high temperature water-cooled divertor for a fusion power reactor

Energy Technology Data Exchange (ETDEWEB)

Giancarli, L. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France)]. E-mail: luciano.giancarli@cea.fr; Bonal, J.P. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France); Puma, A. Li [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France); Michel, B. [CEA Cadarache, Direction de l' Energie Nucleaire, F-13108 St. Paul-les-Durances (France); Sardain, P. [EFDA Close Support Unit, Boltzmannstr. 2, D-85748 Garching (Germany); Salavy, J.F. [CEA Saclay, Direction de l' Energie Nucleaire, F-91191 Gif-sur-Yvette (France)

2005-11-15

This paper presents the conceptual design of a water-cooled divertor target using EUROFER as structural material, water coolant pressure and outlet temperature, respectively, of 15.5 MPa and 325 {sup o}C, and W-alloy monoblocks as armour. Assuming an advanced interface, formed by a thermal barrier in the pipe front part and a compliance layer between W and steel, this concept is able to withstand an incident surface heat flux of 15 MW/m{sup 2}. Both thermal barrier and compliance layer are made of carbon-based materials. The main issues are the manufacturing process of the steel/W interface, and the behaviour under irradiation of graphite materials.

Draining and drying process development of the Tokamak Cooling Water System of ITER

Energy Technology Data Exchange (ETDEWEB)

Kim, Seokho, E-mail: kims@ornl.gov [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Van Hove, Walter; Ferrada, Juan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States); Di Maio, Pietro Alessandro [University of Palermo, Viale delle Scienze, Palermo 90128 (Italy); Felde, David [Reactor and Nuclear Systems Division, ORNL, Oak Ridge, TN (United States); Raphael, Mitteau; Dell’Orco, Giovanni [ITER Organization, 13067 St Paul Lez Durance (France); Berry, Jan [US ITER, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

2016-11-01

Highlights: • A thermal-hydraulic model using RELAP was developed for the ITER FW/BLK modules to determine design parameters for the nitrogen blowout flow rate and pressure. • The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will sufficiently evacuate the water in blankets. • A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation. - Abstract: The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation.

Cleaning device for recycling pump motor cooling system in nuclear reactor

International Nuclear Information System (INIS)

Katayama, Kenjiro; Kondo, Takahisa; Shindo, Kenjiro; Akimoto, Jun.

1996-01-01

The cleaning device of the present invention comprises a cleaning water supply pump, a filter for filtering the cleaning water and a cap member for isolating the inside of a motor casing from the inside of a reactor pressure vessel. A motor in the motor casing and a pump in the reactor pressure vessel are removed, the cap member is attached to the upper end of the motor casing to isolate the inside of the motor casing from the inside of the reactor pressure vessel. If the cleaning water supply pump is operated in this state, the cleaning water flows from a returning pipeline for cooling water circulation, connected to the motor casing to supply pipelines through a heat exchange and is discharged. The discharged water passes through a filter and is sent again, as the cleaning water, to the cleaning water supply pump. With such procedures, the recycling pump motor cooling system in the BWR type reactor can be cleaned without disposing a cyclone separator and irrespective of presence or absence of reactor coolants in the reactor pressure vessel. (I.N.)

A method and programme (BREACH) for predicting the flow distribution in water cooled reactor cores

International Nuclear Information System (INIS)

Randles, J.; Roberts, H.A.

1961-03-01

The method presented here of evaluating the flow rate in individual reactor channels may be applied to any type of water cooled reactor in which boiling occurs The flow distribution is calculated with the aid of a MERCURY autocode programme, BREACH, which is described in detail. This programme computes the steady state longitudinal void distribution and pressure drop in a single channel on the basis of the homogeneous model of two phase flow. (author)

A method and programme (BREACH) for predicting the flow distribution in water cooled reactor cores

Energy Technology Data Exchange (ETDEWEB)

Randles, J; Roberts, H A [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1961-03-15

The method presented here of evaluating the flow rate in individual reactor channels may be applied to any type of water cooled reactor in which boiling occurs The flow distribution is calculated with the aid of a MERCURY autocode programme, BREACH, which is described in detail. This programme computes the steady state longitudinal void distribution and pressure drop in a single channel on the basis of the homogeneous model of two phase flow. (author)

Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

International Nuclear Information System (INIS)

Lee, J.H.; Oka, Y.; Koshizuka, S.

1999-01-01

The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics

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.

Thermal and hydraulic analyses of TFTR cooling water system and magnetic field coils

International Nuclear Information System (INIS)

Lee, A.Y.

1975-10-01

The TFTR toroidal field coils, ohmic heating, hybrid and equilibrium field coils are cooled by water from the machine area cooling water system. The system has the following major equipment and capacities: flow rate of 3600 gpm; ballast tank volume of 5500 gal; pumps of 70.4 m head; chiller refrigeration rating of 3300 tons and connecting pipe of 45.7 cm I.D. The performance of the closed loop system was analyzed and found to be adequate for the thermal loads. The field coils were analyzed with detailed thermal and hydraulic models, including a simulation of the complete water cooling loop. Under the nominal operating mode of one second of toroidal field flat top time and 300 seconds of pulse cycle time, the maximum temperature for the TF coils is 53 0 C; for the OH coils 46 0 C and for the EF coils 39 0 C, which are well below the coil design limit of 120 0 C. The maximum TF coil coolant temperature is 33 0 C which is below the coolant design limit of 100 0 C. The overall pressure loss of the system is below 6.89 x 10 5 Pa (100 psi). With the given chiller refrigeration capacity, the TF coils can be operated to yield up to 4 seconds of flat top time. The TF coils can be operated on a steady state basis at up to 20% of the pulsed duty design current rating of 7.32 kA/coil

Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

International Nuclear Information System (INIS)

Ko, A Reum; Lee, Jeong Ik; Yoon, Ho Joon

2016-01-01

Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases

Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

Energy Technology Data Exchange (ETDEWEB)

Ko, A Reum; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-05-15

Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases.

Cooling system for the IFMIF-EVEDA radiofrequency system

International Nuclear Information System (INIS)

Perez Pichel, G. D.

2012-01-01

The IFMIF-EVEDA project consists on an accelerator prototype that will be installed at Rokkasho (Japan). Through CIEMAT, that is responsible of the development of many systems and components. Empresarios Agrupados get the responsibility of the detailed design of the cooling system for the radiofrequency system (RF system) that must feed the accelerator. the RF water cooling systems is the water primary circuit that provides the required water flow (with a certain temperature, pressure and water quality) and also dissipates the necessary thermal power of all the radiofrequency system equipment. (Author) 4 refs.

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

Pressure drop, steam content and turbulent cross exchange in water/steam flows

International Nuclear Information System (INIS)

Teichel, H.

1978-01-01

For describing the behaviour of two-phase flows of water and steam with the help of calculating patterns, a number of empirical correlations are required. - In this article, correlations for the friction pressure drop in water/steam flows are compared, as well as for the steam mass and the volumetric steam content with each other and with the test results on simple geometries. As the mutual effect between cooling chanels plays an important part at the longitudinal flow through bar bundles, the appertaining equations are evaluated, in addition. (orig.) 891 HP [de

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.

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

TPX heating and cooling system

International Nuclear Information System (INIS)

Kungl, D.J.; Knutson, D.S.; Costello, J.; Stoenescu, S.; Yemin, L.

1995-01-01

TPX, while having primarily super-conducting coils that do not require water cooling, still has very significant water cooling requirements for the plasma heating systems, vacuum vessel, plasma facing components, diagnostics, and ancillary equipment. This is accentuated by the 1000-second pulse requirement. Two major design changes, which have significantly affected the TPX Heating and Cooling System, have been made since the conceptual design review in March of 1993. This paper will discuss these changes and review the current status of the conceptual design. The first change involves replacing the vacuum vessel neutron shielding configuration of lead/glass composite tile by a much simpler and more reliable borated water shield. The second change reduces the operating temperature of the vacuum vessel from 150 C to ≥50 C. With this temperature reduction, all in-vessel components and the vessel will be supplied by coolant at a common ≥50 C inlet temperature. In all, six different heating and cooling supply requirements (temperature, pressure, water quality) for the various TPX components must be met. This paper will detail these requirements and provide an overview of the Heating and Cooling System design while focusing on the ramifications of the TPX changes described above

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.

Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

Science.gov (United States)

Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

2014-08-01

The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

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...

New pool boiling data for water with copper-foam metal at sub-atmospheric pressures: Experiments and correlation

Energy Technology Data Exchange (ETDEWEB)

Choon, Ng Kim; Chakraborty, Anutosh; Aye, Sai Maung; Xiaolin, Wang [Department of Mechanical Engineering, National University of Singapore, 10 Kent Ridge Crescent, Singapore 119260 (Singapore)

2006-08-15

Over the past decades, pool boiling heat transfer of water has been investigated extensively by many scientists and researchers at system pressures varying from atmospheric to near critical pressure. However, at sub-atmospheric pressures conditions there is a dearth of data, particularly when the vapour pressures are less than 10kPa. The authors have conducted a detailed study of pool boiling of water in an evaporator where its system pressure was about 1.8kPa. The heat flux for pool boiling was derived from an uniform radiant heaters up to 5W/cm{sup 2} (or a total heating rate of 125W within an area of 25cm{sup 2}), a region that is of interest for the cooling of CPUs. (author)

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.

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

Solar heating, cooling, and domestic hot water system installed at Kaw Valley State Bank and Trust Company, Topeka, Kansas. Final report

Energy Technology Data Exchange (ETDEWEB)

None

1980-11-01

The building has approximately 5600 square feet of conditioned space. Solar energy is used for space heating, space cooling, and preheating domestic hot water (DHW). The solar energy system has an array of evacuated tube-type collectors with an area of 1068 square feet. A 50/50 solution of ethylene glycol and water is the transfer medium that delivers solar energy to a tube-in-shell heat exchanger that in turn delivers solar-heated water to a 1100 gallon pressurized hot water storage tank. When solar energy is insufficient to satisfy the space heating and/or cooling demand, a natural gas-fired boiler provides auxiliary energy to the fan coil loops and/or the absorption chillers. Extracts from the site files, specification references, drawings, and installation, operation and maintenance instructions are included.

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

Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

International Nuclear Information System (INIS)

Carvalho, F. de A.T. de.

1985-01-01

Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

Operation method and operation control device for emergency core cooling system

Energy Technology Data Exchange (ETDEWEB)

Kinoshita, Shoichiro; Takahashi, Toshiyuki; Fujii, Tadashi [Hitachi Ltd., Tokyo (Japan); Mizutani, Akira

1996-05-07

The present invention provides a method of reducing continuous load capacity of an emergency cooling system of a BWR type reactor and a device reducing a rated capacity of an emergency power source facility. Namely, the emergency core cooling system comprises a first cooling system having a plurality of power source systems based on a plurality of emergency power sources and a second cooling system having a remaining heat removing function. In this case, when the first cooling system is operated the manual starting under a predetermined condition that an external power source loss event should occur, a power source division different from the first cooling system shares the operation to operate the secondary cooling system simultaneously. Further, the first cooling system is constituted as a high pressure reactor core water injection system and the second cooling system is constituted as a remaining heat removing system. With such a constitution, a high pressure reactor core water injection system for manual starting and a remaining heat removing system of different power source division can be operated simultaneously before automatic operation of the emergency core cooling system upon loss of external power source of a nuclear power plant. (I.S.)

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

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)

Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

Energy Technology Data Exchange (ETDEWEB)

Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

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.

Investigation on heat transfer enhancement and pressure loss of double swirl chambers cooling

Directory of Open Access Journals (Sweden)

Gang Lin

2013-09-01

Full Text Available By merging two standard swirl chambers, an alternative cooling configuration named double swirl chambers (DSC has been developed. In the DSC cooling configuration, the main physical phenomena of the swirl flow in swirl chamber and the advantages of swirl flow in heat transfer augmentation are maintained. Additionally, three new physical phenomena can be found in DSC cooling configuration, which result in a further improvement of the heat transfer: (1 impingement effect has been observed, (2 internal heat exchange has been enhanced between fluids in two swirls, and (3 “∞” shape swirl has been generated because of cross effect between two chambers, which improves the mixing of the fluids. Because of all these improvements, the DSC cooling configuration leads to a higher globally-averaged thermal performance parameter (Nu¯¯/Nu∞/(f/f01/3 than standard swirl chamber. In particular, at the inlet region, the augmentation of the heat transfer is nearly 7.5 times larger than the fully developed non-swirl turbulent flow and the circumferentially averaged Nusselt number coefficient is 41% larger than the standard swirl chamber. Within the present work, a further investigation on the DSC cooling configuration has been focused on the influence of geometry parameters e.g. merging ratio of chambers and aspect ratio of inlet duct on the cooling performance. The results show a very large influence of these geometry parameters in heat transfer enhancement and pressure drop ratio. Compared with the basic configuration of DSC cooling, the improved configuration with 20% to 23% merging ratio shows the highest globally-averaged thermal performance parameter. With the same cross section area in tangential inlet ducts, the DSC cooling channel with larger aspect ratio shows larger heat transfer enhancement and at the same time reduced pressure drop ratio, which results in a better globally-averaged thermal performance parameter.

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