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Sample records for heavy water coolant

  1. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  2. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  3. Conceptual designing of reduced-moderation water reactor with heavy water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

    2001-12-01

    The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

  4. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

    2014-01-01

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

  5. Loss of coolant analysis for CIRENE-LATINA heavy water reactor

    International Nuclear Information System (INIS)

    Chiantore, B.; Dubbini, M.; Proto, G.

    1978-01-01

    CIRENE is a heavy-water moderated, boiling water cooled pressure tube reactor. Fuel is natural uranium. A variety of breaks in the primary coolant system have been postulated for the analysis of the CIRENE Latina Plant (now under construction) such as double-end break of inlet header, downcomer, steam line and inlet feeders. The basic tool for analysis is the TILT-N Code which has been purposely developed for simulating the nuclear, thermal and hydrodynamic behaviour of the CIRENE core and associated heat transport system. An extensive full-scale test programme has been carried out by CNEN and CISE which fully confirms the adequacy of the model. The main results of the analysis show that maximum temperatures are far from those leading to significant fuel damage and that adequate core cooling is provided over the whole transient. (author)

  6. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sriram, Jayasree; Vijayan, K.; Kain, Vivekanad; Velmurugan, S.

    2015-01-01

    Advanced Heavy Water Reactor (AHWR) being designed in India proposes to use Plutonium and Thorium as fuel. The objective is to extract energy from the uranium-233 formed from Thorium. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a heavy water moderated and light water cooled tube type boiling water reactor. It is a natural circulation reactor. Thus, it has got several advanced passive safety features built into the system. The various water coolant systems are listed below. i) Main Heat transport System ii) Feed water system iii) Condenser cooling system iv) Process water system and safety systems. As it is a tube type reactor, the radiolysis control differs from the normal boiling water reactor. The coolant enters the bottom of the coolant channel, boiling takes place and then the entire steam water mixture exits the core through the long tail pipes and reaches the moisture separator. Thus, there is a need to devise methods to protect the tail pipes from oxidizing water chemistry condition. Similarly, the moderator heavy water coolant chemistry differs from that of moderator system chemistry of PHWR. The reactivity worth per ppm of gadolinium and boron are low in comparison to PHWR. As a result, much higher concentration of neutron poison has to be added for planned shutdown, start up and for actuating SDS-2. The addition of higher concentration of neutron poison result in higher radiolytic production of deuterium and oxygen. Their recombination back to heavy water has to take into account the higher production of these gases. This paper also discusses the chemistry features of safety systems of AHWR. In addition, the presentation will cover the chemistry monitoring methodology to be implemented in AHWR. (author)

  7. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  8. Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Upadhyaya, B.N.

    2016-01-01

    We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

  9. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

    Purpose: To enable stabilized reactor control, and improve the working rate and the safety of the reactor by removing liquid poison in heavy water while maintaining the power level constant to thereby render the void coefficient of the coolants negative in the low power operation. Method: The operation device for a heavy water moderated reactor comprises a power detector for the reactor, a void coefficient calculator for coolants, control rods inserted into the reactor, a poison regulator for dissolving poisons into or removing them out of heavy water and a device for removing the poisons by the poison regulator device while maintaining the predetermined power level or inserting the control rods by the signals from the power detector and the void coefficient calculator in the high temperature stand-by conditions of the reactor. Then, the heavy water moderated reactor is operated so that liquid poisons in the heavy water are eliminated in the high temperature stand-by condition prior to the start for the power up while maintaining the power level constant and the plurality of control rods are inserted into the reactor core and the void coefficient of the coolants is rendered negative in the low power operation. (Seki, T.)

  10. Tritium concentration in the heavy water upgrading plants

    International Nuclear Information System (INIS)

    Croitoru, C.; Pop, F.; Titescu, Gh.; Dumitrescu, M.; Ciortea, C.; Stefanescu, I.; Peculea, M.; Pitigoi, Gh.; Trancota, D. . E-mail of corresponding author: croitoru@icsi.ro; Croitoru, C.)

    2005-01-01

    In the course of time heavy water used in CANDU nuclear power plants, as moderator or coolant, degrades, as a result of its impurification with light water and tritium. Concentration diminution below 99.8% mol for moderator and 99.75% mol for coolant causes an inefficient functioning of CANDU reactor. By isotopic distillation, light water is removed. Simultaneously tritium concentration takes place. The heavy water upgrading plant from Cernavoda is an isotopic separation cascade with two stages. The paper presents, for this plant, a theoretical study of the tritium concentration. (author)

  11. High conversion heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

    In the present invention, fuel rods using uranium-plutonium oxide mixture fuels are arranged in a square lattice at the same pitch as that in light water cooled reactor and heavy water moderators are used. Accordingly, the volume ratio (Vm/Vf) between the moderator and the fuel can be, for example, of about 2. When heavy water is used for the moderator (coolant), since the moderating effect of heavy water is lower than that of light water, a high conversion ratio of not less than 0.8 can be obtained even if the fuel rod arrangement is equal to that of PWR (Vm/Vf about 2). Accordingly, it is possible to avoid problems caused by dense arrangement of fuel rods as in high conversion rate light water cooled reactors. That is, there are no more troubles in view of thermal hydrodynamic characteristics, re-flooding upon loss of coolant accident, etc., as well as the fuel production cost is not increased. (K.M.)

  12. The heavy water accountancy for research reactors in JAERI

    International Nuclear Information System (INIS)

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

    The three research reactors have been operated by the Department of Research Reactor and used about 41 tons heavy water as coolant, moderator and reflector of research reactors. The JRR-2 is a tank type research reactor of 10MW in thermal power and its is used as moderator, coolant and reflector about 16 tons heavy water. The JRR-3M is a light water cooled and moderated pool type research reactor with a thermal power of 20MW and its is used as reflector about 7.3 tons heavy water. In the JRR-4, which is a light water cooled swimming pool type research reactor with the maximum thermal power of 3.5MW, about 1 ton heavy water is used to supply fully thermalized neutrons with a neutron beam experiment of facility. The heavy water was imported from U.S.A., CANADA and Norway. Parts of heavy water is internationally controlled materials, therefore management of heavy water is necessary for materials accountancy. This report described the change of heavy water inventories in each research reactors, law and regulations for accounting of heavy water in JAERI. (author)

  13. A real-time tritium-in-water monitor for measurement of heavy water leak to the secondary coolant

    International Nuclear Information System (INIS)

    Rathnakaran, M.; Ravetkar, R.M.; Samant, R.K.; Abani, M.C.

    2000-01-01

    The paper describes the development and evaluation of on-line, real-time tritium in water monitor for detection and measurement of heavy water leak to the secondary coolant in a Pressurised Heavy Water Reactor. The detector used for this is a plastic scintillator film, made in the form of sponge and housed in a flow cell which is used for measurement of tritium activity present in heavy water. Two photomultiplier tubes are optically coupled on either face of the flow cell detector and measurement is done in coincidence mode. The sample water is continuously passed through the flow cell detector and a continuous measurement of tritium activity is carried out. It is observed that the impurities in the process water sample are gradually trapped in the flow cell, which affects the transparency of the detector with use. This reduces the sensitivity of the system. In addition, chlorine, which is added in the sample water, to arrest the fungus formation, creates chemiluminescence which interfere the measurement. To improve the sample quality as well as to eliminate the chemiluminescence created by chlorine, sample conditioner consisting of polypropylene candle, activated charcoal and glass fibre filter paper is developed. Polypropylene candle traps particulates above 5 μm pore size, activated charcoal absorbs organic compounds, free chlorine, fungus and turbidity and glass fibre filter paper stops submicron size particles. The measurement is also affected by the interference of dissolved argon-41 in the sample water. A bubbler system developed at BARC is used to strip the dissolved Ar-41 present in the sample which enables the system to measure tritium in presence of this interfering radioactive gas. The microprocessor based electronic system, used in the monitor provides the facility for selection of counting time and thereby improving the counting statistics. Alarm circuit is provided to give timely alarm when the tritium activity concentration exceeds the preset level

  14. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Park, T-K.; Jung, S-H.

    1996-01-01

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  15. Methodologies and technologies for life assessment and management of coolant channels of Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sinha, S.K.; Sinha, R.K.

    2002-01-01

    Zirconium alloy coolant channels are central to the design of Indian Pressurised Heavy Water Reactors (PHWRs) and form the individual pressure boundaries. These coolant channels consist of horizontal pressure tubes made of zirconium alloys, which are separated from cold calandria tubes using garter spring spacers. High temperature heavy water coolant flows through the pressure tube which supports the fuel bundles. A typical coolant channel in a PHWR is shown. These pressure tubes are subjected to several life limiting degradation mechanisms like creep and growth, hydrogen pick-up, reduction in fracture toughness and delayed hydride cracking phenomena because of their operation under high temperature, high stress and high fast neutron flux environment. Considering the early onset of these degradation mechanisms in Zircaloy-2 pressure tubes used in the early generation of Indian PHWRs, the life management of these coolant channels becomes a challenging task, involving multidisciplinary R and D efforts in areas like analytical modelling of degradation mechanisms, evolution of methodologies for assessment of fitness for service and, tools and techniques for remote on line monitoring of integrity, maintenance and replacement. The degradation mechanisms have been modelled and incorporated into specially developed computer codes, such as SCAPCA for irradiation induced creep and growth deformation modelling, HYCON for hydrogen pick-up modelling, BLIST for hydrogen diffusion, blister nucleation and growth modelling and CEAL for assessment of leak before break behaviour. These codes have been validated with respect to the results of in-service inspection and post irradiation examination. Development of analytical models actually paved the way for the evolution of more refined methodologies for assessing the safe residual life of coolant channel. Information gathered from various experiments simulating the degradation mechanisms, results of post-irradiation examination of the

  16. Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2006-01-01

    As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

  17. Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

    2009-01-01

    The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

  18. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  19. Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1991-01-01

    This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

  20. Improvement of lifetime availability through design, inspection, repair and replacement of coolant channels of Indian Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sinha, R.K.

    1998-01-01

    This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

  1. Analysis of actual status of works on technology of heavy liquid metal coolants

    International Nuclear Information System (INIS)

    Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

    2014-01-01

    Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

  2. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

    Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

  3. Parametric studies to establish natural circulation in advanced heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bhatia, S K; Dhawan, M L [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Design of Advanced Heavy Water Reactor (AHWR) is in progress. It consists of vertical pressure tubes with boiling light water coolant flowing through the tubes and heavy water moderator in the calandria. In PHWRs, core heat removal is through forced circulation of the coolant by PHT pumps. In AHWR, no PHT pumps are used and core heat is carried away by natural circulation of the coolant due to density difference between steam/water mixture inside the core and the water region outside the core. This passive means of core heat removal results in a number of benefits viz. (a) extra length of piping, valves, instruments, power supply and control systems for functioning of instruments are eliminated, (b) plant layout is simplified, (c) maintenance of valves and instruments is reduced. Natural circulation in AHWR is achieved by keeping the steam drum at a sufficient height above the core to get the required driving force. The loop height depends on many factors i.e. channel power, V{sub c}/V{sub f} ratio (ratio of coolant volume to fuel volume) and core height. The effect of these parameters on the loop height to establish natural circulation have been studied and presented. (author). 1 ref., 1 fig., 1 tab.

  4. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  5. Development of the heavy-water organic-cooled reactor. Status report from the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Trilling, C A [Atomics International, Division of North American Aviation, Inc., Canoga Park, CA (United States)

    1967-01-01

    In late 1964 the United States Atomic Energy Commission decided to undertake the development of the heavy-water-moderated nuclear power reactor as part of its overall programme for the development of advanced converter reactors. The inclusion of the heavy-water reactor concept was based on its indicated potential for achieving: efficient utilization of available fuel resources; generation of low cost electric power; feasibility of scale-up to very large single unit plant sizes for the dual purpose of generating power and desalting sea water. The excellent neutron economy inherent in heavy-water moderation allows a significant increase in the amount of power which can be generated from a given amount of ore. If one takes into account the amount of uranium required not only for burn-up but also to inventory new reactors in a rapidly expanding nuclear economy, heavy-water reactors show the potential of extracting one and a half to two times more power from the ore mined than light-water reactors. Such an improvement in dynamic fuel utilization will postpone the depletion of low cost uranium ore reserves, providing more time for the discovery of new ore resources and the development of economic fast breeder reactors. The excellent neutron economy of the heavy-water reactor also allows the achievement of appreciable burn-up with low enrichment fuel, with consequent low fuel cycle costs and therefore low energy generation costs. These low fuel cycle costs make the economics of this type of reactor rather insensitive to rising ore costs. They also make the concept well suited for the most economic production of the large quantities of heat required for water desalination. The use of individual pressure tubes for circulating the coolant through the reactor vessel lends itself to the development of a modular type design, which can be scaled up to very large single unit plant sizes by simply increasing the number of identical pressure tube modules and the number of coolant

  6. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

    2001-01-01

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  7. Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

    International Nuclear Information System (INIS)

    Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

    2015-01-01

    At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

  8. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  9. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Kakodkar, A.; Sinha, R.K.; Dhawan, M.L.

    1999-01-01

    Advanced Heavy Water Reactor is a boiling light water cooled, heavy water moderated and vertical pressure tube type reactor with its design optimised for utilisation of thorium for power generation. The core consists of (Th-U 233 )O 2 and (Th-Pu)O 2 fuel with a discharge burn up of 20,000 MWd/Te. This reactor incorporates several features to simplify the design, which eliminate certain systems and components. AHWR design is also optimised for easy replaceability of coolant channels, facilitation of in-service inspection and maintenance and ease of erection. The AHWR design also incorporates several passive systems for performing safety-related functions in the event of an accident. In case of LOCA, emergency coolant is injected through 4 accumulators of 260 m 3 capacity directly into the core. Gravity driven water pool of capacity 6000 m 3 serves to cool the core for 3 days without operator's intervention. Core submergence, passive containment isolation and passive containment cooling are the added features in AHWR. The paper describes the various process systems, core and fuel design, primary components and safety concepts of AHWR. Plant layout and technical data are also presented. The conceptual design of the reactor has been completed, and the detailed design and development is scheduled for completion in the year 2002. (author)

  10. High purity heavy water production: need for total organic carbon determination in process water streams

    International Nuclear Information System (INIS)

    Ayushi; Kumar, Sangita D.; Reddy, A.V.R.; Vithal, G.K.

    2009-01-01

    In recent times, demand for high purity heavy water (99.98% pure) in industries and laboratories has grown by manifold. Its application started in nuclear industry with the design of CANDU reactor, which uses natural uranium as fuel. In this reactor the purest grade of heavy water is used as the moderator and the primary coolant. Diverse industrial applications like fibre optics, medicine, semiconductors etc. use high purity heavy water extensively to achieve better performance of the specific material. In all these applications there is a stringent requirement that the total organic carbon content (TOC) of high purity heavy water should be very low. This is because the presence of TOC can lead to adverse interactions in different applications. To minimize the TOC content in the final product there is a need to monitor and control the TOC content at each and every stage of heavy water production. Hence a simple, rapid and accurate method was developed for the determination of TOC content in process water samples. The paper summarizes the results obtained for the TOC content in the water samples collected from process streams of heavy water production plant. (author)

  11. The key design features of the Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sinha, R.K.; Kakodkar, A.; Anand, A.K.; Venkat Raj, V.; Balakrishnan, K.

    1999-01-01

    The 235 MWe Indian Advanced Heavy Water Reactor (AHWR) is a vertical, pressure tube type, boiling light water cooled reactor. The three key specific features of design of the AHWR, having a large impact on its viability, safety and economics, relate to its reactor physics, coolant channel, and passive safety features. The reactor physics design is tuned for maximising use of thorium based fuel, and achieving a slightly negative void coefficient of reactivity. The fulfilment of these requirements has been possible through use of PuO 2 -ThO 2 MOX, and ThO 2 -U 233 O 2 MOX in different pins of the same fuel cluster, and use of a heterogeneous moderator consisting of pyrolytic carbon and heavy water in 80%-20% volume ratio. The coolant channels of AHWR are designed for easy replaceability of pressure tubes, during normal maintenance shutdowns. The removal of pressure tube along with bottom end-fitting, using rolled joint detachment technology, can be done in AHWR coolant channels without disturbing the top end-fitting, tail pipe and feeder connections, and all other appendages of the coolant channel. The AHWR incorporates several passive safety features. These include core heat removal through natural circulation, direct injection of Emergency Core Coolant System (ECCS) water in fuel, passive systems for containment cooling and isolation, and availability of a large inventory of borated water in overhead Gravity Driven Water Pool (GDWP) to facilitate sustenance of core decay heat removal, ECCS injection, and containment cooling for three days without invoking any active systems or operator action. Incorporation of these features has been done together with considerable design simplifications, and elimination of several reactor grade equipment. A rigorous evaluation of feasibility of AHWR design concept has been completed. The economy enhancing aspects of its key design features are expected to compensate for relative complexity of the thorium fuel cycle activities

  12. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  13. Modeling the transport of hydrogen in the primary coolant of pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Subramanian, H.; Velmurugan, S.; Narasimhan, S.V.; Jain, A.K.; Dash, S.C.

    2008-01-01

    Heavy water (D 2 O) is used in primary heat transport systems of PHWRs. To suppress the radiolysis of heavy water and to control oxygen, hydrogen is added at regular intervals to the primary heat transport system. The added hydrogen finds it way to the heavy water storage tank after passing through the bleed condenser. Owing to the different temperatures and two phase region present in these systems, hydrogen gets redistributed. It is important to know the concentration of dissolved hydrogen in these regions in order to ensure a steady state dissolved hydrogen concentration in the primary system. Different power stations report variations in the frequency and quantity of hydrogen added to achieve the prescribed steady state level. This paper makes an attempt to account for the inventory of hydrogen and model its transport in PHT system. (author)

  14. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  15. Numerical analysis and optimisation of heavy water upgrading column

    International Nuclear Information System (INIS)

    Sankar, Rama; Ghosh, Brindaban; Bhanja, K.

    2013-01-01

    In the 'Pressurised Heavy Water' type of reactors, heavy water is used both as moderator and coolant. During operation of reactor downgraded heavy water is generated that needs to be upgraded for reuse in the reactor. When the isotopic purity of heavy water becomes less than 99.75%, it is termed as downgraded heavy water. Downgraded heavy water also contains impurity such as corrosion products, dirt, oil etc. Upgradation of downgraded heavy water is normally done in two steps: (i) Purification: In this step downgraded heavy water is first purified to remove corrosion products, dirt, oil, etc. and (ii) Upgradation of heavy water to increase its isotopic purity, this step is carried out by vacuum distillation of downgraded heavy water after purification. This project is aimed at mathematical modelling and numerical simulation of heavy water upgrading column. Modelling and simulation studies of the upgradation column are based on equilibrium stage model to evaluate the effect of feed location, pressure, feed composition, reflux ratio in the packed column for given reboiler and condenser duty of distillation column. State to stage modelling of two-phase two-component flow has constitutes the overall modelling of the column. The governing equations consist of stage-wise species and overall mass continuity and stage-wise energy balance. This results in tridigonal matrix equation for stage liquid fractions for heavy and light water. The stage-wise liquid flow rates and temperatures are governed by stage-wise mass and energy balance. The combined form of the corresponding governing equations, with the incorporation of thermodynamic equation of states, form a system of nonlinear equations. This system have been resolved numerically using modified Newton-Raphson method. A code in the MATLAB platform has been developed by on above numerical procedure. The optimisation of the column operating conditions is to be carried out based on parametric studies and analysis of different

  16. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

    1994-12-31

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

  17. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    International Nuclear Information System (INIS)

    Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

    1994-01-01

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

  18. Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

    International Nuclear Information System (INIS)

    Orlov, Yuri I.; Efanov, Alexander D.; Martynov, Pyotr N.; Gulevsky, Valery A.; Papovyants, Albert K.; Levchenko, Yuri D.; Ulyanov, Vladimir V.

    2007-01-01

    In the report, the influence of hydrodynamics of the loop with heavy liquid metal coolants (Pb and Pb-Bi) on the realization methods and efficiency of the coolant technology for the reactor installations of loop, improved loop and mono-block type of design has been studied. The last two types of installations, as a rule, are characterized by the following features: availability of loop sections with low hydraulic head and low coolant velocities, large squares of coolant free surfaces; absence of stop and regulating valve, auxiliary pumps on the coolant pumping-over lines. Because of the different hydrodynamic conditions in the installation types, the tasks of the coolant technology have specific solutions. The description of the following procedures of coolant technology is given in the report: purification by hydrogen (purification using gas mixture containing hydrogen), regulation of dissolved oxygen concentration in coolant, coolant filtrating, control of dissolved oxygen concentration in coolant. It is shown that change of the loop design made with economic purpose and for improvement of the installation safety cause additional requirements to the procedures and apparatuses of the coolant technology realization

  19. Nuclear data needs for subcritical reactors with heavy-metal coolant

    International Nuclear Information System (INIS)

    Ignatyuk, A.V.

    2001-01-01

    Requests on improvement of evaluated data files for minor actinides (MA) are briefly reviewed. New evaluations of neutron cross sections for Np-237, Am-241 and Am-243 after the corresponding tests and verifications should satisfy the required accuracies of data for developing MA-burners. More difficult problems arise for curium isotopes, evaluated data of which are strongly divergent. International expertise of available evaluations could be very desirable. Needs in data improvements for perspective heavy-metal liquid coolants are outlined. (author)

  20. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  1. The condensation of steam on the external surfaces of the shells of HIFAR heavy water heat exchangers during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Chapman, A.G.

    1987-03-01

    A study of steam condensation rates on the HIFAR heavy water heat exchangers was undertaken to predict thermohydraulic conditions in the HIFAR containment during a postulated loss-of-coolant accident (LOCA). The process of surface condensation from a mixture of air and steam, and methods for calculating the rate of condensation, are briefly reviewed. Suitable experimental data are used to estimate coefficients of condensation heat transfer to cool surfaces in a reactor containment during a LOCA. The relevance of the available data to a LOCA in the HIFAR materials testing reactor is examined, and two sets of data are compared. The differences between air/H 2 O and air/D 2 O mixtures are discussed. Formulae are derived for the estimation of the coefficient of heat transfer from the heat exchanger shells to the cooling water, and a method of calculating the rate of condensation per unit area of surface is developed

  2. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    International Nuclear Information System (INIS)

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

  3. Studies on corrosion inhibitors for the cooling water system at the Heavy Water Project, Kota

    International Nuclear Information System (INIS)

    Pillai, B.P.; Mehta, C.T.; Abubacker, K.M.

    1986-01-01

    The Heavy Water Project at Kota uses the water from the Rana Pratap Sagar Lake as coolant in the open recirculation system. In order to find suitable corrosion inhibitors for the above system, a series of laboratory experiments on corrosion inhibitors were carried out using the constructional materials of the cooling water system and a number of proprietary formulations and the results are tabulated. From the data thus generated through various laboratory experiments, the most useful ones have been recommended for application in practice. (author)

  4. Characterization of primary coolant purification system samples for assay of spent ion exchanger radionuclide inventor

    International Nuclear Information System (INIS)

    Sajin Prasad, S.; Pant, Amar; Sharma, Ranjit; Pal, Sanjit

    2018-01-01

    The primary coolant system water of a research reactor contains various fission and activation products and the water is circulated continuously through ion exchange resin cartridges, to reduce the radioactive ionic impurity present in it. The coolant purification system comprises of an ion exchange cooler, two micro filters, and a battery of six ion exchanger beds, associated valves, piping and instrumentation (Heavy water System Operating manual, 2014). The spent cartridge is finally disposed off as active solid waste which contains predominantly long lived fission and activation products. The heavy water coolant is also used to cool the structural assemblies after passing through primary heat exchanger and a metallic strainer, which accumulates the fission and activation products. When there is a reduction of coolant flow through these strainers, they are removed for cleaning and decontamination. This paper describes the characterization of ion exchange resin samples and liquid effluent generated during ultra sonic decontamination of strainer. The results obtained can be used as a methodology for the assay of the spent ion exchanger cartridges radionuclide inventory, during its disposal

  5. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  6. Centrifuge experiments for removal of aluminium turbidity from Dhruva heavy water

    International Nuclear Information System (INIS)

    Shetiya, R.S.; Unny, V.K.P.; Nayak, A.P.

    1989-01-01

    Aluminium turbidity and associated radioactivity was observed in the moderator cum coolant system of Dhruva during initial power operation. Ion exchange resin beds of the purification system were not able to remove aluminium turbidity and radioactivity of system heavy water. Centrifuge technique was used as a convenient alternative method to remove the turbidity and radioactivity. (author)

  7. Various analytical techniques used for the measurement of isotopic purity of heavy water at Madras Atomic Power Station

    International Nuclear Information System (INIS)

    Satyanarayanan, V.; Umapathy, P.; Bhaskaran, R.; Nagarajan, J.; Pradeep, Jeena; Ayyar, S.R.

    2008-01-01

    The paper deals with the various techniques used for the measurement of isotopic purity of heavy water samples received from different sources viz. reactor systems, heavy water upgrading plant and fresh consignment from heavy water production plants. Heavy water is used in PHWRs as moderator and primary coolant. Isotopic Purity is an important parameter to be monitored/analysed regularly for both the systems. There is a minimum isotopic purity level to be maintained in the moderator system due to neutron economy/fuel burnup and in the case of coolant system the measurement is of paramount importance due to its safety considerations. The selection of the method of analysis depends on the isotopic range. The techniques used to measure the isotopic purity of heavy water are a) Infrared Spectrophotometry b) Refractometry c) Densitometry. Infrared spectrometer uses the property of molecular absorption of IR radiation by HOD species and the absorbance is the measure of isotopic purity. This technique is generally used for measuring high isotopic (80-99.98%) and low isotopic samples. Refractometer uses the property of refractive index of heavy water. The difference in refractive indices of light water and heavy water is 0.0048. A 1 % change in D 2 O concentration would thus equal to 0.000048 refractive index units. This method is used for determining the approximate isotopic value of a sample. Density meter uses the property of difference in densities of light and heavy water. The difference in density of 99.999% D 2 O and light water is 0.107540 which covers the whole range of interest. The experience gained with these techniques in the measurements of isotopic purity of various samples are presented in this paper. (author)

  8. Thorium utilization in heavy water moderated Accelerator Driven Systems

    International Nuclear Information System (INIS)

    Bajpai, Anil; Degweker, S.B.; Ghosh, Biplab

    2011-01-01

    Research on Accelerator Driven Systems (ADSs) is being carried out around the world primarily with the objective of waste transmutation. Presently, the volume of waste in India is small and therefore there is little incentive to develop ADS based waste transmutation technology immediately. With limited indigenous U availability and the presence of large Th deposits in the country, there is a clear incentive to develop Th related technologies. India also has vast experience in design, construction and operation of heavy water moderated reactors. Heavy water moderated reactors employing solid Th fuels can be self sustaining, but the discharge burnups are too low to be economical. A possible way to improve the performance such reactors is to use an external neutron source as is done in ADS. This paper discusses our studies on Th utilization in heavy water moderated ADSs. The study is carried out at the lattice level. The time averaged k-infinity of the Th bundle from zero burnup up to the discharge burnup is taken to be the same as the core (ensemble) averaged k-infinity. For the purpose of the analysis we have chosen standard PHWR and AHWR assemblies. Variation of the pitch and coolant (H 2 O/D 2 O) are studied. Both, the once through cycle and the recycling option are studied. In the latter case the study is carried out for various enrichments (% 233 U in Th) of the recycled Th fuel bundles. The code DTF as modified for lattice and burnup calculations (BURNTRAN) was used for carrying out the study. The once through cycle represents the most attractive ADS concept (Th burner ADS) possible for Th utilization. It avoids reprocessing of Th spent fuel and in the ideal situation the use of any fissile material either initially or for sustaining itself. The gain in this system is however rather low requiring a high power accelerator and a substantial fraction of the power generated to be fed back to the accelerator. The self sustaining Th-U cycle in a heavy moderated ADS

  9. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  10. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-01-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  11. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  12. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

    2010-01-01

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  13. Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

    Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

  14. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    International Nuclear Information System (INIS)

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  15. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  16. Detection of heavy-water leaks in nuclear reactors : a novel method

    International Nuclear Information System (INIS)

    Murthy, M.S.; Gor, M.K.

    2002-01-01

    Technical Physics and Prototype Engineering Division, BARC has designed, developed and produced several high sensitivity mass spectrometer helium leak detectors over a period of two decades. Sometimes back, when there was a problem of detecting heavy water leaks in situ in one of the nuclear power reactors of the Department of Atomic Energy, it was referred to this division for a technical solution. After discussing with the site engineers, the various problems involved in the on-line detection of heavy water leaks especially near the end fittings of the coolant assemblies, a novel method of leak detection was developed. Some of the salient features of the method and the results obtained in the laboratory tests are given in this paper. (author)

  17. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  18. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  19. Analysis of loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Moldaschl, H.

    1982-01-01

    Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

  20. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    Science.gov (United States)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  1. Determination of heavy water in heavy water - light water mixtures

    International Nuclear Information System (INIS)

    Sanhueza M, A.

    1986-01-01

    A description about experimental methodology to determine isotopic composition of heavy water - light water mixtures is presented. The employed methods are Nuclear Magnetic Resonance Spectroscopy, for measuring heavy water concentrations from 0 to 100% with intervals of 10% approx., and mass Spectrometry, for measuring heavy water concentrations from 0.1 to 1% with intervals of 0.15% approx., by means of an indirect method of Dilution. (Author)

  2. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  3. Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

    2012-09-01

    In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

  4. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  5. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  6. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  7. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  8. Management of large scale coolant channel replacement programme for Indian PHWRs

    International Nuclear Information System (INIS)

    Bhatnagar, V.K.; Chadda, S.K.; Arya, R.C.

    1994-01-01

    Coolant channel assemblies form most important core components of pressurised heavy water reactors. Zirconium alloy pressure tube which form part of coolant channel assemblies are subjected to environment of high neutron flux, high pressure and temperature. Under those operating environmental conditions, the pressure tubes material undergoes degradation of metallurgical and mechanical properties in addition to dimensional changes. The coolant channels are subjected to an in-service inspection (ISI) programme for monitoring the health particularly of the pressure tubes. The en-mass replacement of pressure tubes is needed after most of the pressure tubes show unacceptable conditions for an assured safe and reliable operation. An overview of various issues pertaining to this aspect is presented. (author). 4 figs

  9. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  10. Heavy water upgrading system in the Fugen heavy water reactor

    International Nuclear Information System (INIS)

    Matsushita, T.; Susaki, S.

    1980-01-01

    The heavy water upgrading system, which is installed in the Fugen heavy water reactor (HWR) was designed to reuse degraded heavy water generated from the deuteration-dedeuteration of resin in the ion exchange column of the moderator purification system. The electrolysis method has been applied in this system on the basis of the predicted generation rate and concentration of degraded heavy water. The structural feature of the electrolytic cell is that it consists of dual cylindrical electrodes, instead of a diaphragm as in the case of conventional water electrolysis. 2 refs

  11. Drying of heavy water system and works of charging heavy water in Fugen

    International Nuclear Information System (INIS)

    Matsushita, Tadashi; Iijima, Setsuo

    1980-01-01

    The advanced thermal reactor ''Fugen'' is the first heavy water-moderated, boiling light water-cooled nuclear reactor for power generation in Japan. It is a large heavy water reactor having about 130 m 3 of heavy water inventory and about 300 m 3 of helium space as the cover gas of the heavy water system. The heavy water required was purchased from FRG, which had been used for the power output test in the KKN, and the quality was 99.82 mol % mean heavy water concentration. The concentration of heavy water for Fugen used for the nuclear design is 99.70 mol%, and it was investigated how heavy water can be charged without lowering the concentration. The matters of investigation include the method of bringing the heavy water and helium system to perfect dryness after washing and light water test, the method of confirming the sufficient dryness to prevent the deterioration, and the method of charging heavy water safely from its containers. On the basis of the results of investigation, the actual works were started. The works of drying the heavy water and helium system by vacuum drying, the works of sampling heavy water and the result of the degree of deterioration, and the works of charging heavy water and the measures to the heavy water remaing in the containers are described. All the works were completed safely and smoothly. (J.P.N.)

  12. Design and development of face seal type sealing plug for advanced heavy water reactor

    International Nuclear Information System (INIS)

    Bansal, S.; Bhattacharyya, S.; Patel, R.J.; Agrawal, R.G.; Vaze, K.K.

    2005-09-01

    Advanced Heavy Water Reactor is a vertical pressure tube type reactor having light water as its coolant and heavy water as moderator. Sealing plug is required to close the pressure boundary of main heat transport system of the reactor by preventing escape of light water/steam From the coolant channel. There are 452 coolant channels in the reactor located in square lattice pitch. Sealing plug is located at the top of each coolant channel (in the top end fitting). Top end fitting is having a stepped bore to create a sealing face. Sealing plug is held through its expanded jaws in a specially provided groove of the end fitting. The plug was designed and prototypes were manufactured considering its functional importance, intricate design and precision machining requirements. Sealing plug consists of about 20 components mostly made up of precipitation hardening stainless steel, which is suitable for water environment and meets other requirements of strength and resistance to wear and galling. Seal disc is a critical component of the sealing plug as it is the pressure-retaining component. It is a circular disc with protruded stem. One face of the seal disc is nickel plated in the peripheral area that creates the sealing by abutting against the sealing face provided in the end fitting. The typical shape and profile of seal disc provides flexibility and allows elastic deformation to assist in locking of sealing plug and creating adequate seating force for effective sealing. Design and development aspects of the sealing plug have been detailed out in this report. Also results of stress analysis and experimental studies for seal disc have been mentioned in the report. Stress analysis and experimental testing was required for the seal disc because high stresses are developed due to its exposure to high pressure and temperature environment of Main Heat Transport system. Hot testing was carried out to simulate the reactor-simulated condition. The performance was found to be

  13. A new sensor for detection of coolant leakage in nuclear power plants using off-axis integrated cavity output spectroscopy

    International Nuclear Information System (INIS)

    Lee, Lim; Park, Hyunmin; Kim, Taek-Soo; Ko, Kwang-Hoon; Jeong, Do-Young

    2012-01-01

    A new sensor based on laser absorption spectroscopy was developed for the detection of coolant leakage which may happen in pressurized heavy water reactor (PHWR). Off-axis integrated output spectroscopy (OA-ICOS) technique was adopted for developing a simple and robust sensor with sufficient sensitivity. Leak events could be monitored by detecting a small change in semi-heavy water (HDO) concentration induced by the exchange reaction of leaked heavy water (D 2 O) with light water (H 2 O). From the results of feasibility tests, we have shown that the measured area of absorption features was linearly correlated with HDO concentration, and the minimum detectable change of HDO concentration with the developed sensor was evaluated as 3.2 ppm. This new sensor is expected to be a reliable and promising device for the detection of coolant leakage since it has some advantages on real-time monitoring and early detection for nuclear safety.

  14. Development of a purification system at Dhruva to treat oil contaminated and chemically impure heavy water

    International Nuclear Information System (INIS)

    Suttraway, S.K.; Mishra, V.; Bitla, S.V.; Ghosh, S.K.

    2006-01-01

    Dhruva, a 100 MW (thermal) Research reactor uses Heavy Water as moderator, reflector and coolant. Normally during plant operation, the Heavy water from the system gets removed during operational and maintenance activities and this collected heavy water gets degraded and contaminated in the process. The degraded heavy water meeting the chemical specification requirement of the up gradation plant is sent for up gradation. Part of the Heavy water collected is contaminated with various organic and inorganic impurities and therefore cannot be sent for IP up gradation as it does not meet the chemical specification of the up gradation plant. This contaminated Heavy water was being stored in SS drums. Over the years of Reactor operation reasonable amount of contaminated Heavy water got collected in the plant. This Heavy water collected from leakages, during routine maintenance, operational activities and fuelling operation had tritium activity and variety of contamination including oil, chlorides, turbidity due to which the specific conductivity was very high. It was decided to purify this Heavy water in house to bring it up to up gradation plant chemical specification requirement. There were number of challenges in formulating a scheme to purify this Heavy water. The scheme needed to be simple and compact in design which could be set up in the plant itself. It should not pose radiological hazards due to radioactive Heavy water during its purification and handling. The contaminated Heavy water collected in drums had varying chemistry and IP. The purification plant should be able to do batch processing so that the different IP and chemical quality of Heavy water stored in different drums are not mixed during purification. It should be capable of removing the oil, chlorides, turbidity and decrease the conductivity to acceptable limits of the Up gradation plant. A purification plant was developed and commissioned after detail laboratory studies and trials. This paper explains

  15. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  16. Hydrodynamics of heavy liquid metal coolant processes and filtering apparatus

    International Nuclear Information System (INIS)

    Albert K Papovyants; Yuri I Orlov; Pyotr N Martynov; Yuri D Boltoev

    2005-01-01

    Full text of publication follows: To optimize the design of filters for cleaning heavy liquid metal coolant (HLMC) from suspended impurities and choose appropriate filter material, the contribution is considered of different mechanisms of delivery and retention of these impurities from the coolant flow, which is governed by its specificity as a thermodynamically instable disperse system to a large extent. It is shown that the buildup of deposits in the filter is favored by the hydrodynamic regime with minimum filtration rates being due to the predominance in the suspension of the fine-dispersed solid phase (oxides Fe 3 O 4 , Cr 2 O 3 and so on). With concentrating the last mentioned phase in filter material pores or stagnant zones, coagulation structuration is possible, which is accompanied by sharp local increase in the viscosity and strength of the solid phase medium being built from liquid metal, i.e. slag sedimentary deposits. In rather extended pores, disintegration of such structures is possible, which is accompanied by sedimentation of large particles produced due to sticking together at coagulation. The analytical solution of the problem of particle sedimentation due to diffusion indicated that in the case under consideration, this mechanism takes place for particles less than ∼ 0,05 μm in size, which is specified by the fact that the time of their delivery to the filter material surface is longer than that of the coolant being in the filter. The London-Van-der-Waals molecular forces play a crucial role in the stage of retention of a separate particle. The constant of the molecular interaction between a spherical particle and the flat surface has been estimated for the chosen value of the gap between the contacting bodies, being dependent on the wetting angle. The sufficient condition for d p -diameter particle capture by the adhesion force field (with a gap of H ≅ 30 nm) is that it be brought by the appropriate forces at a distance from the wall equal

  17. Coolant clean-up system in the primary coolant circuit for nuclear reactor

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

    Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

  18. Determination of total flow rate and flow rate of every operating branch in commissioning of heavy water loop for ARR-2

    International Nuclear Information System (INIS)

    Han Yan

    1997-01-01

    The heavy water loop (i,e, RCS) for ARR-2 in Algeria is a complex loop. Flow regulating means are not provided by the design in order to operate the reactor safely and simplify operating processes. How to determine precisely the orifice diameters of resistance parts for the loop is a key point for decreasing deviation between practical and design flow rates. Commissioning tests shall ensure that under every one of combined operating modes for the pumps, total coolant flow rate is about the same (the number of pumps operating in parallel is the same) and is consistent with design requirement, as well as the distribution of coolant flow rate to every branch is uniform. The flow Determination is divided into two steps. First and foremost, corresponding resistance part at each pump outlet is determined in commissioning test of shorted heavy water loop with light water, so that the problem about uniform distribution of the flow rate to each branch is solved, Secondly, resistance part at the reactor inlet is determined in commissioning test of heavy water loop connected with the vessel, so that the problem about that total heavy water flow rate is within optimal range is solved. According to practical requirements of the project, a computer program of hydraulic calculation and analysis for heavy water loop has been developed, and hydraulic characteristics test for a part of loop has been conducted in order to correct calculation error. By means of program calculation combining with tests in site, orifice diameters of 9 resistance parts has been determined rapidly and precisely and requirements of design and operation has been met adequately

  19. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  20. Deuterium and heavy water

    International Nuclear Information System (INIS)

    Vasaru, G.; Ursu, D.; Mihaila, A.; Szentgyorgyi, P.

    1975-01-01

    This bibliography on deuterium and heavy water contains 3763 references (1932-1974) from 43 sources of information. An author index and a subject index are given. The latter contains a list of 136 subjects, arranged in 13 main topics: abundance of deuterium , catalysts, catalytic exchange, chemical equilibria, chemical kinetics, deuterium and heavy water analysis, deuterium and heavy water properties, deuterium and heavy water separation, exchange reactions, general review, heavy water as moderator, isotope effects, synthesis of deuterium compounds

  1. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  2. Laser-based sensor for a coolant leak detection in a nuclear reactor

    Science.gov (United States)

    Kim, T.-S.; Park, H.; Ko, K.; Lim, G.; Cha, Y.-H.; Han, J.; Jeong, D.-Y.

    2010-08-01

    Currently, the nuclear industry needs strongly a reliable detection system to continuously monitor a coolant leak during a normal operation of reactors for the ensurance of nuclear safety. In this work, we propose a new device for the coolant leak detection based on tunable diode laser spectroscopy (TDLS) by using a compact diode laser. For the feasibility experiment, we established an experimental setup consisted of a near-IR diode laser with a wavelength of about 1392 nm, a home-made multi-pass cell and a sample injection system. The feasibility test was performed for the detection of the heavy water (D2O) leaks which can happen in a pressurized heavy water reactor (PWHR). As a result, the device based on the TDLS is shown to be operated successfully in detecting a HDO molecule, which is generated from the leaked heavy water by an isotope exchange reaction between D2O and H2O. Additionally, it is suggested that the performance of the new device, such as sensitivity and stability, can be improved by adapting a cavity enhanced absorption spectroscopy and a compact DFB diode laser. We presume that this laser-based leak detector has several advantages over the conventional techniques currently employed in the nuclear power plant, such as radiation monitoring, humidity monitoring and FT-IR spectroscopy.

  3. 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)

  4. Heavy water upgrader of 'Fugen'

    International Nuclear Information System (INIS)

    Matsushita, Tadashi; Sasaki, Shigeo

    1980-01-01

    The nuclear power station of the advanced thermal prototype reactor ''Fugen'' has continued the smooth operation since it started the fullscale operation in March, 1979. Fugen is the first heavy water-moderated, boiling light water-cooled reactor in Japan, and its outstanding feature is the use of heavy water as the moderator. The quantity of heavy water retained in Fugen is about 140 m 3 , and the concentration is 99.8 wt.%. This heavy water had been made is USA, and was imported from F.R. of Germany where it had been used. Heavy water is an internationally regulated material, and it is very expensive and hard to purchase. Therefore in order to prevent the deterioration of heavy water and to avoid its loss as far as possible, the management of the quantity and the control of the water quality have been carried out carefully and strictly. The generation of deteriorated heavy water occurs from the exchange of ion exchange resin for poison removal and purification. The heavy water upgrader reconcentrates the deteriorated heavy water of high concentration and returns to the heavy water system, and it was installed for the purpose of reducing the purchase of supplementary heavy water. The outline of the heavy water upgrader, its construction, the performance test and the operation are described. (Kako, I.)

  5. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  6. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  7. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  8. Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

    International Nuclear Information System (INIS)

    Aanstad, O.J.; Sklencar, A.M.

    1983-01-01

    A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

  9. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  10. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  11. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  12. Role of passive valves & devices in poison injection system of advanced heavy water reactor

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

    2014-01-01

    The Advanced Heavy Water Reactor (AHWR) is a 300 MWe pressure tube type boiling light water (H 2 O) cooled, heavy water (D 2 O) moderated reactor. The reactor design is based on well-proven water reactor technologies and incorporates a number of passive safety features such as natural circulation core cooling; direct in-bundle injection of light water coolant during a Loss of Coolant Accident (LOCA) from Advanced Accumulators and Gravity Driven Water Pool by passive means; Passive Decay Heat Removal using Isolation Condensers, Passive Containment Cooling System and Passive Containment Isolation System. In addition to above, there is another passive safety system named as Passive Poison Injection System (PPIS) which is capable of shutting down the reactor for a prolonged time. It is an additional safety system in AHWR to fulfill the shutdown function in the event of failure of wired shutdown systems i.e. primary and secondary shut down systems of the reactor. When demanded, PPIS injects the liquid poison into the moderator by passive means using passive valves and devices. On increase of main heat transport (MHT) system pressure beyond a predetermined value, a set of rupture disks burst, which in-turn actuate the passive valve. The opening of passive valve initiates inrush of high pressure helium gas into poison tanks to push the poison into the moderator system, thereby shutting down the reactor. This paper primarily deals with design and development of Passive Poison Injection System (PPIS) and its passive valves & devices. Recently, a prototype DN 65 size Poison Injection Passive Valve (PIPV) has been developed for AHWR usage and tested rigorously under simulated conditions. The paper will highlight the role of passive valves & devices in PPIS of AHWR. The design concept and test results of passive valves along with rupture disk performance will also be covered. (author)

  13. Design of the solid target structure and the study on the coolant flow distribution in the solid target using the 2-dimensional flow analysis

    International Nuclear Information System (INIS)

    Haga, Katsuhiro; Terada, Atsuhiko; Ishikura, Shuichi; Teshigawara, Makoto; Kinoshita, Hidetaka; Kobayashi, Kaoru; Kaminaga, Masaki; Hino, Ryutaro; Susuki, Akira

    1999-11-01

    A solid target cooled by heavy water is presently under development under the Neutron Science Research Project of the Japan Atomic Energy Research Institute (JAERI). Target plates of several millimeters thickness made of heavy metal are used as the spallation target material and they are put face to face in a row with one to two millimeters gaps in between though which heavy water flows, as the coolant. Based on the design criteria regarding the target plate cooling, the volume percentage of the coolant, and the thermal stress produced in the target plates, we conducted thermal and hydraulic analysis with a one dimensional target plate model. We choosed tungsten as the target material, and decided on various target plate thicknesses. We then calculated the temperature and the thermal stress in the target plates using a two dimensional model, and confirmed the validity of the target plate thicknesses. Based on these analytical results, we proposed a target structure in which forty target plates are divided into six groups and each group is cooled using a single pass of coolant. In order to investigate the relationship between the distribution of the coolant flow, the pressure drop, and the coolant velocity, we conducted a hydraulic analysis using the general purpose hydraulic analysis code. As a result, we realized that an uniform coolant flow distribution can be achieved under a wide range of flow velocity conditions in the target plate cooling channels from 1 m/s to 10 m/s. The pressure drop along the coolant path was 0.09 MPa and 0.17 MPa when the coolant flow velocity was 5 m/s and 7 m/s respectively, which is required to cool the 1.5 MW and 2.5 MW solid targets. (author)

  14. Two-phase coolant pump model of pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Freitas, R.L.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  15. Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

    International Nuclear Information System (INIS)

    Finn, P.A.

    1991-03-01

    Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs

  16. Guidelines to achieve seals with minimal leak rates for HWR-NPR coolant system components

    Energy Technology Data Exchange (ETDEWEB)

    Finn, P.A.

    1991-03-01

    Seal design practices that are acceptable in pressurized-water and boiling-water reactors in the United States are not usable for the Heavy Water Reactor-New Production Reactor (HWR-NPR) because of the stringent requirement on tritium control for the atmosphere within its containment building. To maintain an atmosphere in which workers do not need protective equipment, the components of the coolant system must have a cumulative leak rate less than 0.00026 L/s. Existing technology for seal systems was reviewed with regard to flange, elastomer, valve, and pump design. A technology data base for the designers of the HWR-NPR coolant system was derived from operating experience and seal development work on reactors in the United States, Canada, and Europe. This data base was then used to generate guidelines for the design of seals and/or joints for the HWR-NPR coolant system. Also discussed are needed additional research and development, as well as the necessary component qualification tests for an effective quality control program. 141 refs., 21 figs., 14 tabs.

  17. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  18. Heavy water and nonproliferation

    International Nuclear Information System (INIS)

    Miller, M.M.

    1980-05-01

    This report begins with a historical sketch of heavy water. The report next assesses the nonproliferation implications of the use of heavy water-moderated power reactors; several different reactor types are discussed, but the focus is on the natural uranium, on-power fueled, pressure tube reactor CANDU. The need for and development of on-power fueling safeguards is discussed. Also considered is the use of heavy water in plutonium production reactors as well as the broader issue of the relative nuclear leverage that suppliers can bring to bear on countries with natural uranium-fueled reactors as compared to those using enriched designs. The final chapter reviews heavy water production methods and analyzes the difficulties involved in implementing these on both a large and a small scale. It concludes with an overview of proprietary and nonproliferation constraints on heavy water technology transfer

  19. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  20. Removal of aluminum turbidity from heavy water reactors by precipitation ion exchange using magnesium hydroxide

    International Nuclear Information System (INIS)

    Venkateswarlu, K.S.; Shanker, R.; Velmurugan, S.; Venkateswaran, G.; Rao, M.R.

    1988-01-01

    A special magnesium hydroxide MG(OH)/sub 2/ sorber, loaded onto an ion-exchange matrix has been developed to remove hydrated alumina turbidity in heavy water. This sorber was applied to the coolant/moderator system in the research reactor Dhruva. The sorber not only removed turbidity but also suspended uranium at parts per billion levels and associated β, γ activity. The sorption is based on the attraction between the positively charged Mg(OH)/sub 2/ surface and the negatively charged hydrated alumina particles

  1. Composition and concentration of soluble and particulate matter in the coolant of the reactor primary cooling system of the Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Garcia Rodenas, Luis; La Gamma, Ana M.; Villegas, Marina; Fernandez, Alberto N.; Allemandi, Walter; Manera, Raul; Rosales, Hugo

    2000-01-01

    Nuclear power plants type PWR and PHWR (pressurized water reactor and pressurized heavy water reactor) have three coolant circuits which only exchange energy among them. The primary circuit, whose coolant extracts the reactor energy, the secondary circuit or water-steam cycle and the tertiary circuit which could be lake, river or sea water. The chemistry of the primary and secondary coolants is carefully controlled with the aim of minimizing the corrosion of structural materials. However, very low rates of corrosion are inevitable and one of the consequences of the corrosion processes is the presence of soluble and particulate matter in the coolant from where several problems associated with mass transfer arisen. In this way radioactive nuclides are transported out of the core to the steam generators, hydraulic resistance increases and heat transfer capability degrades. In the present paper some alternative techniques are proposed for the quantification of both, the particulate and soluble matter present in the coolant and their correspondent composition. Some results are also included and discussed. (author)

  2. Breeding capability and void reactivity analysis of heavy-water-cooled thorium reactor

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2008-01-01

    The fuel breeding and void reactivity coefficient of thorium reactors have been investigated using heavy water as coolant for several parametric surveys on moderator-to-fuel ratio (MFR) and burnup. The equilibrium fuel cycle burnup calculation has been performed, which is coupled with the cell calculation for this evaluation. The η of 233 U shows its superiority over other fissile nuclides in the surveyed MFR ranges and always stays higher than 2.1, which indicates that the reactor has a breeding condition for a wide range of MFR. A breeding condition with a burnup comparable to that of a standard PWR or higher can be achieved by adopting a larger pin gap (1-6 mm), and a pin gap of about 2 mm can be used to achieve a breeding ratio (BR) of 1.1. A feasible design region of the reactors, which fulfills the breeding condition and negative void reactivity coefficient, has been found. A heavy-water-cooled PWR-type Th- 233 U fuel reactor can be designed as a breeder reactor with negative void coefficient. (author)

  3. Production of heavy water

    Science.gov (United States)

    Spencer, Larry S.; Brown, Sam W.; Phillips, Michael R.

    2017-06-06

    Disclosed are methods and apparatuses for producing heavy water. In one embodiment, a catalyst is treated with high purity air or a mixture of gaseous nitrogen and oxygen with gaseous deuterium all together flowing over the catalyst to produce the heavy water. In an alternate embodiment, the deuterium is combusted to form the heavy water. In an alternate embodiment, gaseous deuterium and gaseous oxygen is flowed into a fuel cell to produce the heavy water. In various embodiments, the deuterium may be produced by a thermal decomposition and distillation process that involves heating solid lithium deuteride to form liquid lithium deuteride and then extracting the gaseous deuterium from the liquid lithium deuteride.

  4. In-Service Inspection system for coolant channels of Indian PHWRS - evolution and experience

    International Nuclear Information System (INIS)

    Puri, R.K.; Singh, M.

    2006-01-01

    In-Service Inspection (ISI) is the most important of all periodic monitoring and surveillance activities for assuring the structural integrity of coolant channels in the life extension and management of pressurized heavy water reactors (PHWR-CANDU). Indian PHWRs (220 MWe) are characterized by consists by 306 coolant channels in each unit. These channels have to be inspected for various parameters over the operating life of the reactor. ISI of coolant channels necessitated the indigenous development of an inspection system called BARCIS (BARC Channel Inspection System) at Bhabha Atomic Research Center. BARCIS consists of mainly three parts; drive and control unit, special sealing plug and an inspection head carrying various NDT sensors. Five such systems have been built and deployed at various power plants. The paper deals with the development of the BARCIS system for meeting the ISI requirements of coolant channels, development cycle of this system from its conception to evolution to the present state, challenges, data generated and experience gained (ISI of nearly 900 coolant channels has been completed). Prior to BARCIS, pressure tube gauging equipment for pre-service inspection of coolant tubes was developed in 1980. Moreover a tool for ISI of coolant channels in dry condition was developed in 1990. The paper also describes evolution of various contingency procedures and devices developed over the last one decade. Future plans taking into account technological advancement, changes in the scope of inspection due to design and operating experiences and plant layout will also be covered. The paper describes the efforts put in to develop drive and control mechanism to suit the different vault layouts. The drive mechanism is responsible for linear and rotary movement of the inspection head to carry out 100% volumetric inspection. Special emphasis has been laid on the safety devices required during the inspection activity. Special measures for heavy water retention in

  5. Coolant processing device for nuclear reactor

    International Nuclear Information System (INIS)

    Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

    1981-01-01

    Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

  6. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  7. The steam generating heavy water reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

  8. Removal of decay heat by specially designed isolation condensers for advanced heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dhawan, M L; Bhatia, S K [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    For Advanced Heavy Water Reactor (AHWR), removal of decay heat and containment heat is being considered by passive means. For this, special type of isolation condensers are designed. Isolation condensers when submerged in a pool of water, are the best choice because condensation of high temperature steam is an extremely efficient heat transfer mechanism. By the use of isolation condensers, not only heat is removed but also pressure and temperature of the system are automatically controlled without losing the coolant and without using conventional safety relief valves. In this paper, design optimisation studies of isolation condensers of different types with natural circulation for the removal of core decay heat for AHWR is presented. (author). 8 refs., 2 figs.

  9. Analysis of a water-coolant leak into a very high-temperature vitrification chamber

    International Nuclear Information System (INIS)

    Felicione, F. S.

    1998-01-01

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed

  10. Hydrogen radiolytic production in light and heavy water mixtures under conditions similar to LOCA (loss of coolant accidents)

    International Nuclear Information System (INIS)

    Garcia Rodenas, L.; Ali, S.P.; Liberman, S.J.

    1987-01-01

    H 2 , HD and D 2 radiolytic yield in heavy and light water mixtures has been determined to supply the necessary data which will allow to make a realistic estimation of the solution of such gas under LOCA conditions as a function of time. (Author)

  11. 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)

  12. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  13. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  14. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  15. Mid-IR absorption sensing of heavy water using a silicon-on-sapphire waveguide.

    Science.gov (United States)

    Singh, Neetesh; Casas-Bedoya, Alvaro; Hudson, Darren D; Read, Andrew; Mägi, Eric; Eggleton, Benjamin J

    2016-12-15

    We demonstrate a compact silicon-on-sapphire (SOS) strip waveguide sensor for mid-IR absorption spectroscopy. This device can be used for gas and liquid sensing, especially to detect chemically similar molecules and precisely characterize extremely absorptive liquids that are difficult to detect by conventional infrared transmission techniques. We reliably measure concentrations up to 0.25% of heavy water (D2O) in a D2O-H2O mixture at its maximum absorption band at around 4 μm. This complementary metal-oxide-semiconductor (CMOS) compatible SOS D2O sensor is promising for applications such as measuring body fat content or detection of coolant leakage in nuclear reactors.

  16. Heavy Water Quality Management in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Chul; Lee, Mun; Kim, Hi Gon; Park, Chan Young; Choi, Ho Young; Hur, Soon Ock; Ahn, Guk Hoon

    2008-12-15

    Heavy water quality management in the reflector tank is a very important element to maintain the good thermal neutron flux and to ensure the performance of reflector cooling system. This report is written to provide a guidance for the future by describing the history of the heavy water quality management during HANARO operation. The heavy water quality in the reflector tank has been managed by measuring the electrical conductivity at the inlet and outlet of the ion exchanger and by measuring pH of the heavy water. In this report, the heavy water quality management activities performed in HANARO from 1996 to 2007 ere described including a basic theory of the heavy water quality management, exchanging history of used resin in the reflector cooling system, measurement data of the pH and the electrical conductivity, and operation history of the reflector cooling system.

  17. The effect of coolant quantity on local fuel–coolant interactions in a molten pool

    International Nuclear Information System (INIS)

    Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

    2015-01-01

    Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

  18. Production of heavy water in India

    International Nuclear Information System (INIS)

    Deshpande, P.G.; Bimbhat, K.S.; Bhargava, R.K.

    India's first heavy water plant, using electrolysis of water followed by liquid hydrogen distillation, has been operating in association with a fertilizer plant at Nangal since 1962. A dual-temperature process plant at Kota uses heat from the Rajasthan Atomic Power Station. The heavy water plants at Baroda and Tuticorin use ammonia-hydrogen exchange and are integrated with fertilizer ammonia plants. Choice of a particular process for heavy water production depends upon local conditions as well as the extent of the heavy water requirement

  19. Conceptual design of a quasi-homogeneous pressurized heavy water reactor to be operated in the closed Th-U233 fuel cycle

    International Nuclear Information System (INIS)

    1979-06-01

    This paper deals with the heavy water reactor, which, from the neutron economy point of view, offers advantages over the light water reactor. Its capability to be fuelled with natural uranium has also been considered a desirable nuclear option by various countries with sufficient domestic uranium resources not wishing to be dependent on the import of enrichment and other fuel cycle services which, in addition, would draw on the foreign exchange reserves. Pressurized heavy water reactors have been designed and built according to two somewhat different versions. While the Canadian CANDU-PHWR concept uses pressure tubes in a nearly unpressurized moderator tank (calandria), the German development line takes advantage of the established and well proven LWR technology, and, thus, uses a pressure vessel design where coolant channels and the surrounding moderator are held at equal pressure. This pressure vessel type heavy water reactor which has been built on a commercial demonstration plant level at ATUCHA in Argentina is described in a companion paper where also a conceptual design for a 685 MWsub(e) PHWR is discussed

  20. Canadian heavy water production

    International Nuclear Information System (INIS)

    Dahlinger, A.; Lockerby, W.E.; Rae, H.K.

    1977-05-01

    The paper reviews Canadian experience in the production of heavy water, presents a long-term supply projection, relates this projection to the anticipated long-term electrical energy demand, and highlights principal areas for further improvement that form the bulk of our research and development program on heavy water processes

  1. Power control device for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Matsushima, Hidesuke; Masuda, Hiroyuki.

    1978-01-01

    Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

  2. A simple and rapid gas chromatographic method for the determination of dissolved deuterium and nitrogen in heavy water coolant of a nuclear reactor

    International Nuclear Information System (INIS)

    Nair, B.K.S.

    1976-01-01

    A known volume of a heavy water sample is equilibrated with a known volume of pure helium gas at atmospheric pressure in a sample tube. The dissolved gases evolve into the helium and distribute themselves between the gaseous and liquid phases according to their equilibrium partial pressures. These partial pressures of the gases in the equilibrium gas mixture are determined by analysing it gas-chromatographically. From these analytical data and the absorption coefficients of deuterium and nitrogen, their original concentrations in heavy water are calculated. Corrections for the increase in the total pressure of the gaseous phase owing to evolved gases are calculated and found to be negligible. Air contamination during sampling and analysis can be detected by the presence of the oxygen peak in the chromatogram and corrected for. The calculation is facilitated by programming it on an electronic calculator. The method is much simpler and faster than the vacuum method usually applied for this analysis. One determination can be completed in about an hour. The average deviation and standard deviation have been estimated at 0.19 ml/litre heavy water and 0.25 ml/litre heavy water respectively in deuterium, and 0.36 and 0.68 ml/litre in nitrogen. (author)

  3. MIC damage in a water coolant header for remote process equipment

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1996-01-01

    Stainless steel water piping, used to supply coolant for remote chemical separations equipment, developed several leaks during low flow conditions, the result of an extended interruption of operations. All the leaks occurred at welds in the bottom of the pipe, which was blanketed with silt deposits from unfiltered well water used for cooling. Ultrasonic, radiographic, and metallographic examinations of the leak sites revealed worm-hole pitting adjacent to the welds. Seepage at the penetrations was strongly acidic and corroded the external pipe surfaces. Analyses of the water and deposits suggested microbiologically influenced corrosion and fouling

  4. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  5. Preliminary assessment of water-based nano-fluids for use as coolants in PWRs

    International Nuclear Information System (INIS)

    Jacopo Buongiorno

    2005-01-01

    Full text of publication follows: The impact of using water-based fluids with small additions (<2% vol.) of nano-sized (10-100 nm) particle populations as coolants for current and advanced PWRs is evaluated. Such 'engineered' fluids (known as nano-fluids) are attractive because the presence of the nano-particles enhances energy transport considerably. As a result, nano-fluids are known to have (i) higher thermal conductivity than water (up to 20% depending on nano-particle material, size and volumetric fraction), (ii) higher heat transfer coefficients (up to 40%), (iii) higher CHF (up to 300% in pool boiling), and (iv) comparable pressure drop. Furthermore, nano-fluids appear to be very stable suspensions with little or no sedimentation, because of the small size of the dispersed particles and their typically low volumetric fractions. The ultimate objective of this work is to assess whether existing PWRs could be retro-fitted with a water-based nano-fluid coolant, to increase safety margins, reduce stored energy, and/or allow for power up-rates. Also, advanced PWRs could be designed with nano-fluids. The linear heat generation rate in PWRs is limited by a) fuel centerline melting, b) cladding overheating (CHF), and c) stored energy release following a large-break LOCA. Mechanisms b) and c) are usually the most limiting. For given geometry and linear power, it is obvious that the core with the nano-fluid coolant will have higher margins to CHF and LOCA limits. Conversely, for given margins, a higher linear power can be accommodated by the nano-fluid-cooled core. Standard thermal-hydraulic models for the PWR hot fuel pin (including a RELAP model for the LOCA) have been used to quantify the benefit of using nano-fluid coolants on the performance of a PWR. (author)

  6. All heavy metals closed-cycle analysis on water-cooled reactors of uranium and thorium fuel cycle systems

    International Nuclear Information System (INIS)

    Permana, Sidik; Sekimoto, Hiroshi; Waris, Abdul; Takaki, Naoyuki

    2009-01-01

    Uranium and Thorium fuels as the basis fuel of nuclear energy utilization has been used for several reactor types which produce trans-uranium or trans-thorium as 'by product' nuclear reaction with higher mass number and the remaining uranium and thorium fuels. The utilization of recycled spent fuel as world wide concerns are spent fuel of uranium and plutonium and in some cases using recycled minor actinide (MA). Those fuel schemes are used for improving an optimum nuclear fuel utilization as well to reduce the radioactive waste from spent fuels. A closed-cycle analysis of all heavy metals on water-cooled cases for both uranium and thorium fuel cycles has been investigated to evaluate the criticality condition, breeding performances, uranium or thorium utilization capability and void reactivity condition. Water-cooled reactor is used for the basic design study including light water and heavy water-cooled as an established technology as well as commercialized nuclear technologies. A developed coupling code of equilibrium fuel cycle burnup code and cell calculation of SRAC code are used for optimization analysis with JENDL 3.3 as nuclear data library. An equilibrium burnup calculation is adopted for estimating an equilibrium state condition of nuclide composition and cell calculation is performed for calculating microscopic neutron cross-sections and fluxes in relation to the effect of different fuel compositions, different fuel pin types and moderation ratios. The sensitivity analysis such as criticality, breeding performance, and void reactivity are strongly depends on moderation ratio and each fuel case has its trend as a function of moderation ratio. Heavy water coolant shows better breeding performance compared with light water coolant, however, it obtains less negative or more positive void reactivity. Equilibrium nuclide compositions are also evaluated to show the production of main nuclides and also to analyze the isotopic composition pattern especially

  7. Symposium on operational and environmental issues concerning use of water as a coolant in power plants and industries: proceedings

    International Nuclear Information System (INIS)

    2008-12-01

    The symposium is organised to bring together researchers, plant operators and regulatory agencies working in the area of operational and environmental problems associated with use of water as a coolant in power plants and other allied industries. The symposium targets chemists, biologists, environmental scientists, power plant operating engineers and plant designers working in various academic, governmental and non-governmental organisations. The major themes of the symposium are: water chemistry of coolant systems in power plants and other industries, chemistry of primary and moderator systems in nuclear power plants and research reactors, corrosion issues including Flow-Accelerated Corrosion (FAC) and its control in water coolant systems, chemistry of steam and water at elevated temperature in nuclear power plants, once through steam generator chemistry, industrial fire water systems, ion-exchange purification, innovative water treatment in power and industrial units, chemical cleaning and chemical decontamination, biofouling and biocorrosion, cooling water treatment chemicals and their environmental fate and environmental impact of thermal effluents. Papers relevant to INIS are indexed separately

  8. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  9. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  10. Technical solutions for tritium removal from Cernavoda NPP heavy water systems

    International Nuclear Information System (INIS)

    Barariu, Gheorghe; Panait, Adrian

    2002-01-01

    In CANDU nuclear plants 2400 KCi/GW(e) - year tritium is generated. At a CANDU - 600 reactor similar to Cernavoda NPP Unit 1, 1500 KCi/year of tritium is generated 95% being in the D 2 O moderator, which can achieve a radioactivity level of 80 - 100 Ci/kg. Tritium in heavy water contributes with 30 - 50% to the doses received by operation personnel and with 20% to the radioactivity released to the environment. The extraction of tritium heavy water at CANDU reactors implies the following possibilities: - the radioactivity level reduction in the operation area; - the maintenance and repair cost reduction due to reduction of personnel protection measures and increased labor productivity; - the increase of NPP utilization factor by shutdown time reduction for maintenance and repair; - tritium concentration reduction from technological systems, ensuring thus the possibility of redesigning the systems in order to lower the cost of investment; - profitable use of extracted tritium. Technical measures provided by AECL project for CANDU 600 at Cernavoda make possible to satisfy the current standards concerning tritium concentration in the operation area atmosphere of 5 x 10 -6 Ci/m 3 . The regulations recommend that the radioactivity level should be maintained as low as possible in conformity with ALARA principles. Also, it is possible that norms will become more restrictive in the future, so the tritium removal technology is a good preventive measure which may become very necessary. The methods, which currently reached the industrial or pilot stages, are based on catalyzed chemical exchange, the heavy water electrolysis, and deuterium distillation. They are known as: VPCE - Vapour Phase Catalytic Exchange; LPCE - Liquid Phase Catalytic Exchange; DE - Direct Electrolysis; CD - Cryogenic Distillation. As transfer processes the catalyzed chemical exchange and heavy water electrolysis are used while concentration of tritium gas is done by cryogenic distillation. At present the

  11. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2001-01-01

    The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

  12. Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-02-15

    When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

  13. Method of controlling power of a heavy water reactor

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1975-01-01

    Object: To adjust a level of heavy water in a region of reflection body to control power in a heavy water reactor. Structure: The interior of a core tank filled with heavy water is divided by a partition into a core heavy water region and a reflection body region formed by surrounding the core heavy water region, and a level of heavy water within the reflection body region is adjusted to control power. Preferably, it is desirable to communicate the core heavy water region with the reflection body heavy water region at their lower portion, and gas pressure applied to an upper portion within at least one of said regions is adjusted to adjust the level of heavy water within the reflection body heavy water region. Thereby, the heavy water within the reflection body heavy water region may be introduced into the core region, thus requiring no tank which stores heavy water within the reflection body region. (Kamimura, M.)

  14. Measurement of concentration of heavy water

    International Nuclear Information System (INIS)

    Tsukamoto, Yuichi; Kondo, Mitsuo; Sakurai, Naoyuki

    1979-01-01

    The concentration of heavy water is measured as one of the technical management in the Fugen plant. The heavy water is used as the moderator in the reactor. The measuring method depends on the theory of light absorption. The light absorption range of heavy water spreads from near infrared to infrared zone. The near infrared absorption was adopted for the purpose, as the absorption is much larger in infrared zone, and the measurement has to be conducted, limiting the apparent absorption. This measuring method is available to determine the concentration of heavy water in the broad range exactly. The preparation of heavy water sample and the measurement of the absorption spectra of near infrared ray are explained, as the experimental procedure. The sample cell was made of quartz, and the spectroscope was the Hitachi 323 type. The resolving power is 100 nm and 27 nm for the wave length of 1000 nm and 2500 nm, respectively. Concerning the measured results, the absorption was recorded in the wave length range from 600 nm to 2600 nm, and for the heavy water concentration range from 0 to 99.77 wt. %. The peaks of absorption were located at the wave length of 1450, 1660, 1920, 1970, 2020 and 2600 nm. The three kinds of fundamental vibration mode of the molecules of both light and heavy water are shown, and the peaks belong to H 2 O, HDO and D 2 O, respectively. The relation between the absorption and the heavy water concentration, and that between the transmissivity and the wave length are shown, when the cell thickness was varied to 5 mm and 20 mm, and the heavy water concentration to 21%, 62% and 99.85%. (Nakai, Y.)

  15. Design of a molten heavy-metal coolant and target for fast-thermal accelerator driven sub-critical system (ADS)

    International Nuclear Information System (INIS)

    Satyamurthy, P.; Degwekar, S.B.; Nema, P.K.

    2001-01-01

    Accelerator Driven sub-critical Systems (ADS) have evoked considerable interest in recent years. The Energy Amplifier concept developed by C. Rubbia and others at CERN incorporates a buoyancy driven, lead-coolant primary system for extracting the heat generated in the fast reactor as well as that in neutron spallation target. In earlier publications, our BARC group has proposed a one-way coupled booster reactor system which could be operated at proton beam currents as low as 1-2 mA for a power output of 750 MW th . Here, the basic idea is to have a fast booster reactor zone of low power (- 100 MW th ) which is separated by a large gap from the main thermal reactor zone. In this arrangement, the spallation neutron source feeds neutrons to the fast reactor zone where neutrons are further multiplied. Further in this system, the neutrons from the booster region enter the main reactor but very few neutrons from main reactor return to booster, thus ensuring one-way coupling. In earlier work, several possible configurations of the booster and thermal regions were presented. In the present work, we describe an engineering design particularly with respect to thermal hydraulics of lead/lead-bismuth eutectic coolant also acting as spallation neutron source. This hybrid ADS reactor consists of fast and thermal reactor zones producing about 100 MW th and 650 MW th respectively. The scheme of the system is shown. The fast core consists of 48 hexagonal fuel bundles each containing 169 fuel pins of 8.2 mm diameter arranged in 11.4 mm triangular array pitch. The average thermal power per fuel pin is about 13.46 kw. However, due to neutron flux peaking effect, the maximum fuel pin power can be up to 2.5 times this average power. The thermal reactor consists of heavy water as moderator and coolant similar to a typical CANDU type Indian PHWR except for fuel composition. Though the gap between fast and thermal zones essentially provides one way coupling of neutron flux, a thermal

  16. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    Science.gov (United States)

    Zapata, Luis E.; Hackel, Lloyd

    1999-01-01

    A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

  17. Knock-limited performance of several internal coolants

    Science.gov (United States)

    Bellman, Donald R; Evvard, John C

    1945-01-01

    The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

  18. Study of the heavy water regeneration processes

    International Nuclear Information System (INIS)

    Cavcic, E.

    1965-11-01

    Experience derived from heavy water reactor operation showed degradation and dilution of heavy water to be inevitable and depends on the type of reactor. Dilution of heavy water during operation of the RA and the RB reactors is shown in this report. Principles and procedures of heavy water regeneration by electrolysis, fractional distillation, cleaning, prevention of tritium contamination are described as well as separation columns

  19. Heavy Water - Industrial Separation Processes

    International Nuclear Information System (INIS)

    Peculea, M.

    1984-01-01

    This monograph devoted to the heavy water production mainly presents the Romanian experience in the field which started in early sixties from the laboratory scale production and reached now the level of large scale industrial production at ROMAG-Drobeta, Romania. The book is structured in eleven chapters entitled: Overview, The main physical properties, Sources, Uses, Separation factor and equilibrium constant, Mathematical modelling of the separation process, Thermodynamical considerations on the isotope separation, Selection criteria for heavy water separation processes, Industrial installations for heavy water production, Prospects, Acknowledgements. 200 Figs., 90 Tabs., 135 Refs

  20. Role of research and development in life management programme and upgradation of safety of Indian Pressurised Heavy Water Reactors

    International Nuclear Information System (INIS)

    Rupani, B.B.; Vijayan, P.K.; Rama Rao, A.; Sinha, R.K.

    2009-01-01

    At present, India has a fleet of thirteen small size 220 MWe Pressurised Heavy Water Reactors (PHWRs) and two medium size 540 MWe PHWRs. Reactor Engineering Division (RED) of Bhabha Atomic Research Centre (BARC) has pursued multi-faceted Research and Development programmes to support each phase of PHWR i.e. design, construction, commissioning, operation, maintenance, In-Service Inspection, repair and replacement and life extension, This programme is mainly related to life management of coolant channels, development of tooling and techniques for In-service Inspection of coolant channels, development of repair and replacement technology for coolant channels and moderator system, In-house development of technology and equipments like rolled joints to joint dissimilar metals and lancing equipment for steam generator and state-of art diagnostic systems for trouble shooting critical operating systems. The strong R and D support provided in the programme has significantly contributed towards safe operation of PHWRs. This paper gives the highlights of the major activities in above areas with their end uses and capability. (author)

  1. Coolant system decontamination

    International Nuclear Information System (INIS)

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

    1981-01-01

    An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

  2. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  3. Transient simulation of coolant peak temperature due to prolonged fan and/or water pump operation after the vehicle is keyed-off

    Science.gov (United States)

    Pang, Suh Chyn; Masjuki, Haji Hassan; Kalam, Md. Abul; Hazrat, Md. Ali

    2014-01-01

    Automotive designers should design a robust engine cooling system which works well in both normal and severe driving conditions. When vehicles are keyed-off suddenly after some distance of hill-climbing driving, the coolant temperature tends to increase drastically. This is because heat soak in the engine could not be transferred away in a timely manner, as both the water pump and cooling fan stop working after the vehicle is keyed-off. In this research, we aimed to visualize the coolant temperature trend over time before and after the vehicles were keyed-off. In order to prevent coolant temperature from exceeding its boiling point and jeopardizing engine life, a numerical model was further tested with prolonged fan and/or water pump operation after keying-off. One dimensional thermal-fluid simulation was exploited to model the vehicle's cooling system. The behaviour of engine heat, air flow, and coolant flow over time were varied to observe the corresponding transient coolant temperatures. The robustness of this model was proven by validation with industry field test data. The numerical results provided sensible insights into the proposed solution. In short, prolonging fan operation for 500 s and prolonging both fan and water pump operation for 300 s could reduce coolant peak temperature efficiently. The physical implementation plan and benefits yielded from implementation of the electrical fan and electrical water pump are discussed.

  4. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

    2012-07-01

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

  5. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  6. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  7. Finishing and upgrading of heavy water

    International Nuclear Information System (INIS)

    Butler, J.P.; Hammerli, M.

    1981-01-01

    This invention provides a process and apparatus for deuterium enrichment as a final stage in a heavy water plant, for continuous on-line enrichment of the heavy water in moderator and heat transfer systems in heavy water nuclear reactors, and for enrichment of hevy water that has been downgraded with natural water during the course of operating a heavy water nuclear reactor. The method comprises contacting partially-enriched heavy water feed in a catalyst column with hydrogen gas (essentially D 2 ) orginating in an electrolysis cell so as to enrich the feed water with deuterium extracted from the electrolytic hydrogen gas and passing the deuterium-enriched water to the electrolysis cell. The apparatus comprises a catalyst isotope exchange column with hydrogen gas and liquid water passing through in countercurrent isotope exchange, an electrolysis cell, a dehumidifer-scrubber; and means for passing the liquid water enriched in deuterium from the catalyst column through the dehumidifer-scrubber to the electrolysis cell, for passing the hydrogen gas evolved in the cathode side of the cell through the dehumidifier-scrubber to the catalyst column, for passing the hydrogen gas from the catalyst column to an output, for introducing an input water feed to the upper portion of the catalyst column, and for taking a product enriched in deuterium from the system. (LL)

  8. Measurement of tritium activity in the aluminum pipe of JRR-2 heavy water primary cooling system using imaging plate

    International Nuclear Information System (INIS)

    Motoishi, Shoji; Kobayashi, Katsutoshi

    2000-12-01

    JRR-2 is the heavy water cooling type nuclear reactor, which has been operated for 36 years (1960-1976) and in the process of decommissioning at present. For this reason, evaluation of tritium quantity permeated into the pipe and apparatus of the primary coolant heavy water circulating system is important. In the Radioisotope Production Division, activity of tritium in aluminum pipe was measured with imaging plate (IP), liquid scintillation analyzer and high purity germanium detector (HPGe). After acrylic paints was applied for the region except for tritium contamination on the surface of aluminum pipe, only the oxidized contaminated part was dissolved by 1.5%(1.21M) HF for 3 minutes, and measured with IP. As a result, the tritium was found to permeate in the depth of 25 μm. Moreover, 90% of it was found to be distributed within 7 μm. (author)

  9. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  10. Factors governing particulate corrosion product adhesion to surfaces in water reactor coolant circuits

    International Nuclear Information System (INIS)

    1979-03-01

    Gravity, van der Waals, magnetic, electrical double layer and hydrodynamic forces are considered as potential contributors to the adhesion of particulate corrosion products to surfaces in water reactor coolant circuits. These forces are renewed and evaluated, and the following are amongst the conclusions drawn; adequate theories are available to estimate the forces governing corrosion product particle adhesion to surfaces in single phase flow in water reactor coolant circuits. Some uncertainty is introduced by the geometry of real particle-surface systems. The major uncertainties are due to inadequate data on the Hamaker constant and the zeta potential for the relevant materials, water chemistry and radiation chemistry at 300 0 C; van der Waals force is dominant over the effect of gravity for particles smaller than about 100 m; quite modest zeta potentials, approximately 50mV, are capable of inhibiting particle deposition throughout the size range relevant to water reactors; for surfaces exposed to typical water reactor flow conditions, particles smaller than approximately 1 m will be stable against resuspension in the absence of electrical double layer repulsion; and the magnitude of the electrical double layer repulsion for a given potential depends on whether the interaction is assumed to occur at constant potential or constant change. (author)

  11. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  12. MIC damage in a water coolant header for remote process equipment

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1994-01-01

    Stainless steel water piping used to supply coolant for remote chemical separations equipment developed leaks during low flow conditions resulting from an extended interruption of operations. All the leaks occurred at welds in the bottom zone of the pipe, which was blanketed with silt deposits from the unfiltered well water used for cooling. Ultrasonic, radiographic, and metallographic examinations of leak sites revealed worm hole pitting adjacent to the welds. Seepage at the penetrations was strongly acidic and resulted in corrosion on the external pipe surfaces beneath brown crusty deposits which had developed. Analyses of the water and deposits suggest a strong propensity toward microbiologically influenced corrosion (MIC) and fouling

  13. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  14. Effect of ionite decomposition products on the reactor coolant pH in a boiling-water reactor

    International Nuclear Information System (INIS)

    Bredikhin, V.Ya.; Moskvin, L.N.

    1982-01-01

    The effect of products resulting from thermal radiolysis of ionites on water-chemical regime of NPP with RBMK is considered basing on investigations conducted in a boiling type experimental reactor. Data are presented on dynamics of changes in the specific electric conductivity and pH of the coolant following destruction of ion exchange groups and ionite matrix under the effect of reactor radiation. The authors draw a conclusion that radiation destruction of ionito fine disperse suspension or high-molecular soluble compounds in the reactor are, probably, one of the main reasons for variations in pH values of the coolant at NPP in non-correction water chemical regime

  15. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  16. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  17. Heavy water at Aswan

    International Nuclear Information System (INIS)

    1959-01-01

    A fertilizer factory is being built by Egyptian Chemical Industries (Kima) at Aswan on the upper Nile; it will produce a mixture of ammonium nitrate and calcium carbonate adjusted to contain 20.5% nitrogen. It is also proposed to construct a heavy water plant to be located at and integrated with the fertilizer factory. At the request of the Government of the United Arab Republic, the International Atomic Energy Agency sent an expert to carry out investigation of the technical, economic and other related aspects of the proposed production of heavy water. A report was submitted to the IAEA Director General. Its main conclusions can be summarized as follows: (1) Production of heavy water as a by-product of fertilizer manufacture at Aswan is technically feasible. Separation of deuterium from industrial hydrogen for this purpose could be done either by catalytic exchange or by liquefaction and distillation; the choice should depend on economic considerations. (2) The heavy water produced at Aswan should be competitive in cost with that produced elsewhere; this, however, would depend on whether firm contracts are obtained for the delivery of equipment at guaranteed prices and with guaranteed performance, and whether such prices are in reasonable agreement with preliminary estimates. (3) The future market for heavy water is difficult to predict. For one thing, there is a very large production capacity in the USA, most of which is idle due to lack of demand. Secondly, there is a relatively small production outside the USA that is sold at prices higher than that charged by the US Government. The future of the market is necessarily contingent upon the possibility of future free sale by the US Government. At the end of his report, the expert has also given his comments on possible further assistance to the project by IAEA

  18. Indian heavy water programme - challenges and opportunities

    International Nuclear Information System (INIS)

    Aruldoss Kanthiah, W.S.

    2010-01-01

    Discovery of fission of uranium in 1939 opened up hitherto unknown possibilities for utilising the fission energy for use of mankind, mainly for the production of and electrical energy. It was realised that this nuclear energy could be an ideal substitute for the fast depleting fossil fuels which would one day get exhausted. Two main concepts of nuclear power reactor got evolved, one enriched uranium fuelled, ordinary water moderated reactor and another natural uranium fuelled heavy water moderated reactor. The concentration of uranium 235 U needed for ordinary water moderated reactors is 3% but the naturally occurring uranium in India contains only 0.7% of 235 U. The reactors utilising natural uranium as fuel require Heavy Water as moderator. The processing of uranium ore to achieve from 0.7% to 3% is highly complex. Recognising the fact that India has limited uranium resources but rich thorium resources, Dr. Bhabha formulated a three stage nuclear power generation programme for our country. The first generation reactors can use natural uranium as fuel with heavy water as moderator. Since the technology to generate such large scale heavy water to match the urgent need for nuclear power generation was not indigenously available, the technology available with Canada and France was utilised for installation of first generation heavy water plants in India. However, the peaceful nuclear experiment conducted by India in 1974 caused resentment among the countries that supplied Heavy Water technology to India and they stopped all technological help and assistance in nuclear field. Thereafter, it was the story of India going alone in heavy water production. That made India meets successfully all challenges on the way to installation, commissioning and sustained operation of all plants. Today we have six operating Heavy Water plants, spread all over the country. We have reached a stage, a change from a situation of crunch to a level of not only self sufficiency but to a

  19. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

  20. Technical status study of heavy water enrichment

    International Nuclear Information System (INIS)

    Sukarsono; Imam Dahroni; Didik Herhady

    2007-01-01

    Technical status study of heavy water enrichment in Indonesia and also in the world has been done. Heavy water enrichment processes have been investigated were water distillation, hydrogen distillation, laser enrichment, electrolysis and isotop exchange. For the isotop exchange, the chemical pair can be used were water-hydrogen sulphite, ammonium-hydrogen, aminomethane-hydrogen, and water-hydrogen. For the isotope exchange, there was carried out by mono thermal or bi thermal. The highest producer of heavy water is Canada, and the other producer is USA, Norwegian and India. The processes be used in the world are isotope exchange Girdler Sulphide (GS), distillation and electrolysis. Research of heavy water carried out in Batan Yogyakarta, has a purpose to know the characteristic of heavy water purification. Several apparatus which has erected were 3 distillation column: Pyrex glass of 2 m tall, stainless steel column of 3 m tall and steel of 6 m tall. Electrolysis apparatus is 50 cell electrolysis and an isotope exchange unit which has catalyst: Ni- Cr 2 O 3 and Pt-Carbon. These apparatus were not ready to operate. (author)

  1. Canadian heavy water production

    International Nuclear Information System (INIS)

    Dahlinger, A.; Lockerby, W.E.; Rae, H.K.

    1977-01-01

    The paper reviews Canadian experience in the production of heavy water, presents a long-term supply projection, relates this projection to the anticipated long-term electrical energy demand, and highlights principal areas for further improvement that form the bulk of the Canadian R and D programme on heavy water processes. Six Canadian heavy water plants with a total design capacity of 4000Mg/a are in operation or under construction. All use the Girdler-Sulphide (GS) process, which is based on deuterium exchange between water and hydrogen sulphide. Early operating problems have been overcome and the plants have demonstrated annual capacity factors in excess of 70%, with short-term production rates equal to design rates. Areas for further improvement are: to increase production rates by optimizing the control of foaming to give both higher sieve tray efficiency and higher flow rates, to reduce the incapacity due to deposition of pyrite (FeS 2 ) and sulphur (between 5% and 10%), and to improve process control and optimization of operating conditions by the application of mathematical simulations of the detailed deuterium profile throughout each plant. Other processes being studied, which look potentially attractive are the hydrogen-water exchange and the hydrogen-amine exchange. Even if they become successful competitors to the GS process, the latter is likely to remain the dominant production method for the next 10-20 years. This programme, when related to the long-term electricity demand, indicates that heavy water supply and demand are in reasonable balance and that the Candu programme will not be inhibited because of shortages of this commodity. (author)

  2. Topical and working papers on heavy water requirements and availability

    International Nuclear Information System (INIS)

    The documents included in this report are: Heavy water requirements and availability; technological infrastructure for heavy water plants; heavy water plant siting; hydrogen and methane availability; economics of heavy water production; monothermal, water fed heavy water process based on the ammonia/hydrogen isotopic exchange; production strategies to meet demand projections; hydrogen availability; deuterium sources; the independent UHDE heavy water process

  3. Considerations regarding design of ion exchange columns for applications in heavy water nuclear reactors- a comprehensive review

    International Nuclear Information System (INIS)

    Joginder Kumar; Nema, M.K.

    2000-01-01

    In nuclear reactor applications the principal role of the purification system is to maintain a satisfactory chemistry of moderator and coolant which are different at various stages of reactor operations e.g. during reactor start up, for removal of neutron poison from the moderator, the purification flows are much different compared to steady state operation of the reactor. In order to cater to varying requirements regarding purification load, optimisation in connection with ion exchange column design plays an important role and becomes very challenging in Heavy Water Nuclear Reactors mainly due to the fact that heavy water is very very expensive. In this paper a comprehensive review is made for various designs adopted so far regarding IX column in Indian PHWRs of 220 MWe size for normal operations. Design and operating experience regarding large size IX column used for occasional needs during dilute chemical decontamination of 220 MWe PHWRs is also discussed. The experience regarding development testing of the proposed design of ion exchange column for 500 MWe PHWRs is also discussed

  4. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

    Guthrie, J.E.; Acres, O.E.

    1979-03-01

    Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

  5. Independent modification on water lubrication loop of radial-axial bearing of Russian reactor coolant pump

    International Nuclear Information System (INIS)

    Gu Yingbin

    2012-01-01

    Water lubrication was used for radial-axial bearings of 1391M reactor coolant pumps at both units of Tianwan Nuclear Power Plant Phase I Project, which was the first trial on large commercial pressurized water reactors in the world. As a prototype, there were inherent deficiencies leading to a series of operational events. Jiangsu Nuclear Power Corporation conducted the independent innovative technical modification to cope with the defects, and succeeded in reducing heat removal rate of the radial-axial bearings of the reactor coolant pumps, mitigating or preventing the cavitation abrasion of the bearings and improving the cooling effects. This paper illustrates the reasons of the innovative modification, the design and implementation preparation of modification program, the implementation process and evaluation of modification effect, including detailed follow-up work program. (author)

  6. Infrared thermal measurements of laser soft tissue ablation as a function of air/water coolant for Nd:YAG and diode lasers

    Science.gov (United States)

    Gekelman, Diana; Yamamoto, Andrew; Oto, Marvin G.; White, Joel M.

    2003-06-01

    The purpose of this investigation was to measure the maximum temperature at the Nd:YAG and Diode lasers fiberoptic tips as a function of air/water coolant, during soft tissue ablation in pig jaws. A pulsed Nd:YAG laser (1064nm) and a Diode laser (800-830 nm) were used varying parameters of power, conditioning or not of the fiber tip, under 4 settings of air/water coolant. The maximum temperature at the fiber tip was measured using an infra-red camera and the interaction of the fiber with the porcine soft tissue was evaluated. A two-factor ANOVA was used for statistical analysis (plaser interaction with soft tissues produced temperatures levels directly proportional to power increase, but the conditioning of the fiber tip did not influence the temperature rise. On the other hand, conditioning of the fiber tip did influence the temperature rise for Diode laser. The addition of air/water coolant, for both lasers, did not promote temperature rise consistent with cutting and coagulation of porcine soft tissue. Laser parameters affect the fiberoptic surface temperature, and the addition of air/water coolant significantly lowered surface temperature on the fiberoptic tip for all lasers and parameters tested.

  7. Stresses imposed by coolant channel end shield interaction in 200 MWe PHWR

    International Nuclear Information System (INIS)

    Mehra, V.K.; Singh, R.K.; Soni, R.S.; Kushwaha, H.S.; Kakodkar, A.

    1983-01-01

    End shield of 200 MWe Pressurised Heavy Water Reactor (PHWR) is a composite tube sheet structure consisting of two circular tube sheets joined together by lattice tubes. Each lattice tube houses a coolant channel assembly which is connected to the end shield through shock absorber device. End shield assembly is suspended in the vault by hanger rods and its horizontal position is controlled by a set of pre-compressed springs. Coolant channel assemblies elongate due to their exposure to fast neutron flux in the reactor. This permanent elongation is monitored periodically. When growth of the channel exceeds a present value, it is prevented from further elongation by the shock absorbing device. Resultant force exerted on the end shield makes it move. This paper describes a numerical method used for evaluating these forces and movement of the end shield. Stresses produced by these forces are calculated by using finite element method. Typical stress values are verified by strain gauge measurements. (orig.)

  8. Definition and Analysis of Heavy Water Reactor Benchmarks for Testing New Wims-D Libraries

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2000-01-01

    This work is part of the IAEA-WIMS Library Update Project (WLUP). A group of heavy water reactor benchmarks have been selected for testing new WIMS-D libraries, including calculations with WIMSD5B program and the analysis of results.These benchmarks cover a wide variety of reactors and conditions, from fresh fuels to high burnup, and from natural to enriched uranium.Besides, each benchmark includes variations in lattice pitch and in coolants (normally heavy water and void).Multiplication factors with critical experimental bucklings and other parameters are calculated and compared with experimental reference values.The WIMS libraries used for the calculations were generated with basic data from JEF-2.2 Rev.3 (JEF) and ENDF/B-VI iNReleaseln 5 (E6) Results obtained with WIMS-86 (W86) library, included with WIMSD5B package, from Windfrith, UK with adjusted data, are included also, for showing the improvements obtained with the new -not adjusted- libraries.The calculations with WIMSD5B were made with two methods (input program options): PIJ (two-dimension collision probability method) and DSN (one-dimension Sn method, with homogenization of materials by ring).The general conclusions are: the library based on JEF data and the DSN meted give the best results, that in average are acceptable

  9. Flooding of a large, passive, pressure-tube light water reactor

    International Nuclear Information System (INIS)

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

    1997-01-01

    A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

  10. The Canadian heavy water situation

    International Nuclear Information System (INIS)

    Dahlinger, A.

    1975-08-01

    Existing heavy water plants in Canada are producing at a satisfactory rate and currently planned capacity is in balance with projected needs. By 1980, we shall have Girdler-Sulphide plants installed with a design capacity of almost 600 kg/h. Effort is required to minimize production costs for heavy water and to ensure that costs do not increase faster than the current inflationary trend. (Author)

  11. The purification of organic reactor coolants; La purification des refrigerants organiques

    Energy Technology Data Exchange (ETDEWEB)

    Hannaert, H; Lopes Cardozo, R [CCR EURATOM, Ispra, Varese (Italy)

    1967-01-01

    Among the main impurities present in irradiated and virgin coolants we have been particularly interested in chlorine, water, iron, oxygen and heavy elements. Our studies have been directed along two basic lines, namely: (1) the elimination of inorganic impurities, and (2) the elimination of organic impurities. The purpose of the studies on the elimination of inorganic impurities is to obtain a 'clean' coolant (virgin or used), this cleanness being marked by a low tendency to form deposits on the fuel element cladding (fouling corrosion). Careful attention has been paid to the problems of chlorine, water, iron and coated particles. Particular interest has been attached to research on deoxygenation of the organic liquid by catalytic hydrocracking, the oxygen (which comes from the dissociation of the water and the gases contained in the liquid) favouring polymerization and the formation of particles which are likely to be deposited on the hot walls. As regards the elimination of degradation products, many studies have been carried out with a view to permitting the maximum recycling of decomposed hydrocarbides and thus reducing the cost of make-up, the recycling rate being a function of the mean molecular weight and the viscosity of the recycled fractions. The technique based on solvent extraction has led to very satisfactory results. An extension of this technique, using counter-flow extraction, appears to be even more promising. (author)

  12. IR analyzer spots heavy water leaks

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    A correlation spectrometer developed by Barringer Research Ltd. (in collaboration with Atomic Energy of Canada and Ontario Hydro) is used to measure HDO concentrations in DTO in the final (distillation) stage of heavy-water production. A unit has been installed at Bruce Heavy Water Plant. Previously, such spectrometers had been installed to detect heavy-water leaks in CANDU reactors. The principle on which the instrument works is explained, with illustrations. It works by comparing the absorption at 2.9 μm, due to HDO, with that at 2.6 μm, due to both HDO and D 2 O. (N.D.H.)

  13. Moderator clean-up system in a heavy water reactor

    International Nuclear Information System (INIS)

    Sasada, Yasuhiro; Hamamura, Kenji.

    1983-01-01

    Purpose: To decrease the fluctuation of the poison concentration in heavy water moderator due to a heavy water clean-up system. Constitution: To a calandria tank filled with heavy water as poison-containing moderators, are connected both end of a pipeway through which heavy water flows and to which a clean-up device is provided. Strongly basic resin is filled within the clean-up device and a cooler is disposed to a pipeway at the upstream of the clean-up device. In this structure, the temperature of heavy water at the inlet of the clean-up device at a constant level between the temperature at the exit of the cooler and the lowest temperature for the moderator to thereby decrease the fluctuation in the poison concentration in the heavy water moderator due to the heavy water clean-up device. (Moriyama, K.)

  14. Future trends in heavy water production

    International Nuclear Information System (INIS)

    Galley, M.R.

    1983-10-01

    World heavy water production has spanned nearly fifty years and, for much of that period, the commodity was often in short supply, but that situation has changed, at least in Canada. There are now adequate reserves of heavy water and sufficient installed production capacity to service Canadian domestic and export demands for the next ten years or beyond. More than 90 percent of the world's inventory of heavy water has been produced by the GS process but this may not be the method that is chosen when the time comes to expand heavy water production again. Other countries, such as India and Argentina, have already chosen ammonia-hydrogen exchange as an alternative technology for part of their domestic production programs. Despite the present surplus of heavy water, research and development of new technologies is very active, particularly in Canada and Japan, because it is recognized that there are still attractive opportunities for future production by processes that are both less expensive and environmentally more acceptable, than either the demonstrated GS process or ammonia-hydrogen alternative. This paper describes the prospects for some of these new processes, contrasts them with the present established methods and assesses the probable impact on the future supply situation

  15. Water issues associated with heavy oil production.

    Energy Technology Data Exchange (ETDEWEB)

    Veil, J. A.; Quinn, J. J.; Environmental Science Division

    2008-11-28

    Crude oil occurs in many different forms throughout the world. An important characteristic of crude oil that affects the ease with which it can be produced is its density and viscosity. Lighter crude oil typically can be produced more easily and at lower cost than heavier crude oil. Historically, much of the nation's oil supply came from domestic or international light or medium crude oil sources. California's extensive heavy oil production for more than a century is a notable exception. Oil and gas companies are actively looking toward heavier crude oil sources to help meet demands and to take advantage of large heavy oil reserves located in North and South America. Heavy oil includes very viscous oil resources like those found in some fields in California and Venezuela, oil shale, and tar sands (called oil sands in Canada). These are described in more detail in the next chapter. Water is integrally associated with conventional oil production. Produced water is the largest byproduct associated with oil production. The cost of managing large volumes of produced water is an important component of the overall cost of producing oil. Most mature oil fields rely on injected water to maintain formation pressure during production. The processes involved with heavy oil production often require external water supplies for steam generation, washing, and other steps. While some heavy oil processes generate produced water, others generate different types of industrial wastewater. Management and disposition of the wastewater presents challenges and costs for the operators. This report describes water requirements relating to heavy oil production and potential sources for that water. The report also describes how water is used and the resulting water quality impacts associated with heavy oil production.

  16. Ultrasonic relaxation studies associated with n-octylamine-heavy water

    International Nuclear Information System (INIS)

    Kor, S.K.; Singh, R.K.

    1994-01-01

    Ultrasonic absorption measurements have been carried out in lyotropic liquid crystalline system n-octylamine/heavy water in the frequency range 5-65 MHz and at temperatures 30 degC and 37 degC at different concentrations of heavy water in octylamine. Velocity has been measured using interferometric technique at 2 MHz at different concentrations of heavy water. Ultrasonic absorption coefficients at different concentrations in the concentration range 0.3-0.9 m.f. heavy water have been found to show a maxima in the absorption curve at critical concentration (∼0.85 m.f. heavy water). This peak has been found to shift towards lower concentrations of heavy water at higher frequencies. Results have been analysed and it has been found that single relaxation process takes place around 30 MHz and this has been attributed to aggregation of octylamine and heavy water molecules. (author). 8 refs., 6 figs., 3 tabs

  17. A study of Cirus heavy water system isotopic purity

    International Nuclear Information System (INIS)

    Thomas, Shibu; Sahu, A.K.; Unni, V.K.P.; Pant, R.C.

    2000-01-01

    Cirus uses heavy water as moderator and helium as cover gas. Approximately one tonne of heavy water was added to the system every year for routine make up. Isotopic purity (IP) of this water used for addition was always higher than that of the system. Though this should increase IP of heavy water in the system, it has remained almost at the same level, over the years. A study was carried out to estimate the extent of improvement in IP of heavy water in the system that should have occurred because of this and other factors in last 30 years. Reasons for non-occurrence of such an improvement were explored. Ion exchange resins used for purification of heavy water and air ingress into helium cover gas system appear to be the principal sources of entry of light water into heavy water system. (author)

  18. Radiation safety experience in upgrading 2-5% heavy water wastes at Heavy Water Plant, Nangal (Preprint No. SA-7)

    International Nuclear Information System (INIS)

    Sadhukhan, H.K.; Behl, D.; Ramraj; Iyengar, T.S.; Sadarangani, S.H.; Vaze, P.K.; Soman, S.D.

    1989-04-01

    This paper describes the radiological safety experience in upgrading 2-5% heavy water wastes at Heavy Water Plant at Nangal at the third stage electrolysers. The feed water concentrations at the third stage electrolyer was determined after a safety analysis study and pilot plant experiment, which gave the optimal concentrations of 1 to 1.5 mCi (3.7 to 5.5 x 10 7 Bq) per litre per minute feed from a submerged SS tank containing 2-5% heavy water wastes. This process not only yielded an efficient recovery of reactor grade heavy water but contained the tritium activity in the third stage electrolysers and in the final product viz., heavy water. The tritium concentrations were continuously monitore d by liquid scintillation counting method at all the three stages of electrolysis plant, the distillation plant, the heavy water filling rooms, the drains, the ambient air, the product fertilizer (calcium ammonia nitrate) and the Sutlej River and found to be well within the safety limits set for general public at large. The HD and D 2 process streams in the palnt were monitored using fill-in type of ionization chambers designed for the purpose, which served a D 2 inventory check as well. There was no internal exposure to any personnel during the entire period of programme. (author). 2 tabs

  19. Economic and safety aspects of using moderator heat for feed water heating in a nuclear power plant

    International Nuclear Information System (INIS)

    Patwegar, I.A.; Dutta, Anu; Chaki, S.K.; Venkat Raj, V.

    2002-01-01

    Full text: In the proposed advanced heavy water reactor (AHWR), coolant and moderator are separated by the coolant channel. The coolant absorbs most of the fission heat produced in the reactor core. However, the moderator absorbs about 5 to 6 % of the fission heat. In a reactor producing 750 MW(th) power, this moderator heat is about 40 MW. In the present Indian PHWR (pressurized heavy water reactor) systems, this moderator heat is lost to a sink through the moderator heat exchangers, which are cooled by process water. This paper presents the results of the steam cycle analysis carried out for AHWR using moderator heat exchangers as part of the feed heating system. The present study is an attempt to determine the gain in electrical output (MW) if moderator heat is utilized for feed water heating. The operational and safety aspects of using moderator heat are also discussed in the paper

  20. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

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

  1. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  2. Heavy water leak detection using diffusion sampler

    International Nuclear Information System (INIS)

    Joshi, M.L.; Hussain, S.A.

    1990-01-01

    In the Pressurrised Heavy Water Reactors (PHWRs) detection of the sources of heavy water leaks is importent both for the purpose of radiation hazard control as well as for the reduction of escape/loss of heavy water which, is an expensive nuclear material. This paper describes an application of tritium diffusion sampler for heavy water leak detection. The diffusion sampler comprises an usual tritium counting glass vial with a special orifice. The counting vial has water vapour, deficient in HTO concentration. The HTO present outside diffuses in the vial through the orifice, gets exchanged with water of the wet filter paper kept at the bottom and the moisture in the vial atmosphere which has HTO concentration lower than that outside. This results in continuation of net movement of HTO in the vial. The exchanged tritium is counted in liquid scintillation spectrometer. The method has a sensitivity of 10000 dpm/DAC-h. (author). 2 figs., 2 ta bs

  3. Topical and working papers on heavy water accountability and safeguards

    International Nuclear Information System (INIS)

    This report contains the following papers: 1) Statement of IAEA concerning safeguarding of heavy water; 2) Preliminary Canadian Comments on IAEA document on heavy water safeguards; 3) Heavy water accountability 03.10.78; 4) Heavy water accountability 05.04.79

  4. Limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.

    1985-01-01

    The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

  5. Composition and concentration of soluble and particulate matter in the coolant of the reactor primary cooling system of the Embalse nuclear power plant; Composicion y concentracion del material soluble y particulado en el refrigerante del SPTC de la central nuclear Embalse

    Energy Technology Data Exchange (ETDEWEB)

    Chocron, Mauricio; Garcia Rodenas, Luis; La Gamma, Ana M; Villegas, Marina [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Quimica; Fernandez, Alberto N; Allemandi, Walter; Manera, Raul; Rosales, Hugo [Nucleoelectrica Argentina SA (NASA), Embalse (Argentina). Central Nuclear Embalse

    2000-07-01

    Nuclear power plants type PWR and PHWR (pressurized water reactor and pressurized heavy water reactor) have three coolant circuits which only exchange energy among them. The primary circuit, whose coolant extracts the reactor energy, the secondary circuit or water-steam cycle and the tertiary circuit which could be lake, river or sea water. The chemistry of the primary and secondary coolants is carefully controlled with the aim of minimizing the corrosion of structural materials. However, very low rates of corrosion are inevitable and one of the consequences of the corrosion processes is the presence of soluble and particulate matter in the coolant from where several problems associated with mass transfer arisen. In this way radioactive nuclides are transported out of the core to the steam generators, hydraulic resistance increases and heat transfer capability degrades. In the present paper some alternative techniques are proposed for the quantification of both, the particulate and soluble matter present in the coolant and their correspondent composition. Some results are also included and discussed. (author)

  6. Plutonium Recycle Test Reactor (PRTR). Operating Experience and Supporting R and D, Its Application to Heavy-Water Power Reactor Design and Operation

    Energy Technology Data Exchange (ETDEWEB)

    Harty, H. [Battelle Memorial Institute, Pacific Northwest Laboratories, Richland, WA (United States)

    1968-04-15

    Convincing answers to questions about heavy-water, pressure-tube, power reactors, e.g. pressure-tube serviceability, heavy-water management problems, long-term behaviour of special pressure-tube reactor components, and unique operating maintenance problems (compared to light-water reactors) must be based on actual operating experience with that type of reactor. PRTR operating experience and supporting R and D studies, although not always simple extrapolations to power reactors, can be summarized in a context applicable to future heavy-water power reactors, as follows: 1. Pressure-tube life, in a practical case, need not be limited by creep, gross hydriding, corrosion, or mechanical damage. The possibility that growth of a defect (perhaps service-induced) to a size that is critical under certain operating conditions, remains a primary unknown in pressure- tube life extrapolations. A pressure-tube failure in PRTR (combined with gross release of fuel material) proved only slightly more inconvenient, time consuming, and damaging to the reactor proper, than occurred with a gross failure of a fuel element in PRTR. 2. Routine operating losses of heavy water appear tractable in heavy-water-cooled power reactors; losses from low-pressure systems can be insignificant over the life of a plant. Non-routine losses may prove to be the largest component of loss over the life of a plant. 3. The performance of special components in PRTR, e.g. the calandria and shields, has not deteriorated despite being subjected to non-standard operating conditions. The calandria now contains a light-water reflector with single barrier separation from the heavy-water moderator. The carbon steel shields (containing carbon steel shot) show no deterioration based on pressure drop measurements and piping activation immediately outside the shields. The helium pressurization system (for primary coolant pressurization) remains a high maintenance system, and cannot be recommended for power reactors, based

  7. CFD Application and OpenFOAM on the 2-D Model for the Moderator System of Heavy-Water Reactors

    International Nuclear Information System (INIS)

    Chang, Se Myong; Park, A. Y.; Kim, Hyoung Tae

    2011-01-01

    The flow in the complex pipeline system in a calandria tank of CANDU reactor is transported through the distribution of heat sources, which also exerts the pressure drop to the coolant flow. So the phenomena should be considered as multi-physics both in the viewpoints of heat transfer and fluid dynamics. In this study, we have modeled the calandria tank system as two-dimensional simplified one preliminarily that is yet far from the real objects, but to see the essential physics and to test the possibility of the present CFD(computational fluid dynamics) methods for the thermo-hydraulic problem in the moderator system of heavy-water reactors

  8. Effect of coolant velocity on the fragmentation of single melt drops in water

    International Nuclear Information System (INIS)

    Cunningham, M.H.; Frost, D.L.

    1997-01-01

    Flash X-ray radiography and high-speed photography are used to investigate the effect of the coolant velocity on the fine fragmentation of molten tin drops in water. A water cannot is used to accelerate the water to a constant speed of up to 30 m/s. The water is accelerated with a double piston arrangement including a foam shock absorber to eliminate the formation of a shock wave. In this way, the effect of coolant velocity on drop breakup is investigated in the absence of the strong shock wave that is present in most earlier studies. The results show that there is a transition from thermal to hydrodynamic fragmentation through an intermediate stage in which the drops initially undergo hydrodynamic fragmentation followed by the formation of a vapour bubble. For low velocities (9 m/s) this bubble collapses, fragmenting the remainder of the drop while at greater velocities (15 m/s) the drop breaks up within the bubble before it condenses. At 22 and 28 m/s there is no vapour formation and the drop fragments due to hydrodynamic effects. Quantitative analysis of the radiographs is used to determine the mass distribution of the melt during the drop fragmentation. Comparison with earlier work in which the ambient flow is preceded by a strong shock wave indicates that the transition from thermal to hydrodynamic breakup is strongly dependent on the characteristics of the pressure field experienced by the drop. (author)

  9. Heavy water lattices: Second panel report

    International Nuclear Information System (INIS)

    1963-01-01

    The panel was attended by prominent physicists from most of the laboratories engaged in the field of heavy water lattices throughout the world. The participants presented written contributions and status reports describing the past history and plans for further development of heavy-water reactors. Valuable discussions took place, during which recommendations for future work were formulated. Refs, figs, tabs

  10. Heavy water lattices: Second panel report

    Energy Technology Data Exchange (ETDEWEB)

    1963-09-15

    The panel was attended by prominent physicists from most of the laboratories engaged in the field of heavy water lattices throughout the world. The participants presented written contributions and status reports describing the past history and plans for further development of heavy-water reactors. Valuable discussions took place, during which recommendations for future work were formulated. Refs, figs, tabs.

  11. The Canadian heavy water supply program

    International Nuclear Information System (INIS)

    Dahlinger, A.; McNally, P.J.

    1976-06-01

    The performance to date of individual Canadian heavy water plants is described in detail as are the current plant construction plans. These data, when related to the long-term electricity demand indicate that heavy water supply and demand are in reasonable balance and that the CANDU program will not be inhibited because of shortages of the commodity. (author)

  12. Heavy water. A production alternative for Venezuela

    International Nuclear Information System (INIS)

    A survey of heavy water production methods is made. Main facts about isotopic and distillation methods, reforming and coupling to a Hydrogen distillation plant are presented. A feasibility study on heavy water production in Venezuela is suggested

  13. Zero waste machine coolant management strategy at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Carlson, B.; Algarra, F.; Wilburn, D.

    1998-01-01

    Machine coolants are used in machining equipment including lathes, grinders, saws and drills. The purpose of coolants is to wash away machinery debris in the form of metal fines, lubricate, and disperse heat between the part and the machine tool. An effective coolant prolongs tool life and protects against part rejection, commonly due to scoring or scorching. Traditionally, coolants have a very short effective life in the machine, often times being disposed of as frequently as once per week. The cause of coolant degradation is primarily due to the effects of bacteria, which thrive in the organic rich coolant environment. Bacteria in this environment reproduce at a logarithmic rate, destroying the coolant desirable aspects and causing potential worker health risks associated with the use of biocides to control the bacteria. The strategy described in this paper has effectively controlled bacterial activity without the use of biocides, avoided disposal of a hazardous waste, and has extended coolant life indefinitely. The Machine Coolant Management Strategy employed a combination of filtration, heavy lubricating oil removal, and aeration, which maintained the coolant peak performance without the use of biocides. In FY96, the Laboratory generated and disposed of 19,880 kg of coolants from 9 separate sites at a cost of $145K. The single largest generator was the main machine shop producing an average 14,000 kg annually. However, in FY97, the waste generation for the main machine shop dropped to 4,000 kg after the implementation of the zero waste strategy. It is expected that this value will be further reduced in FY98

  14. A thermal analysis computer programme package for the estimation of KANUPP coolant channel flows and outlet header temperature distribution

    International Nuclear Information System (INIS)

    Siddiqui, M.S.

    1992-06-01

    COFTAN is a computer code for actual estimation of flows and temperatures in the coolant channels of a pressure tube heavy water reactor. The code is being used for Candu type reactor with coolant flowing 208 channels. The simulation model first performs the detailed calculation of flux and power distribution based on two groups diffusion theory treatment on a three dimensional mesh and then channel powers, resulting from the summation of eleven bundle powers in each of the 208 channels, are employed to make actual estimation of coolant flows using channel powers and channel outlet temperature monitored by digital computers. The code by using the design flows in individual channels and applying a correction factor based on control room monitored flows in eight selected channels, can also provide a reserve computational tool of estimating individual channel outlet temperatures, thus providing an alternate arrangements for checking Rads performance. 42 figs. (Orig./A.B.)

  15. Heavy water physical verification in power plants

    International Nuclear Information System (INIS)

    Morsy, S.; Schuricht, V.; Beetle, T.; Szabo, E.

    1986-01-01

    This paper is a report on the Agency experience in verifying heavy water inventories in power plants. The safeguards objectives and goals for such activities are defined in the paper. The heavy water is stratified according to the flow within the power plant, including upgraders. A safeguards scheme based on a combination of records auditing, comparing records and reports, and physical verification has been developed. This scheme has elevated the status of heavy water safeguards to a level comparable to nuclear material safeguards in bulk facilities. It leads to attribute and variable verification of the heavy water inventory in the different system components and in the store. The verification methods include volume and weight determination, sampling and analysis, non-destructive assay (NDA), and criticality check. The analysis of the different measurement methods and their limits of accuracy are discussed in the paper

  16. Seismic re-evaluation of Heavy Water Plant, Kota

    International Nuclear Information System (INIS)

    Parulekar, Y.M.; Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

    2003-10-01

    This report deals with seismic re-evaluation of Heavy Water Plant, Kota. Heavy Water Plant, Kota handles considerable amount of H 2 S gas, which is very toxic. During the original design stage as per IS 1893-1966 seismic coefficient for zone-I was zero. Therefore earthquake and its effects were not considered while designing the heavy water plant structures. However as per IS 1893 (1984) the seismic coefficient for zone-I is 0.01 g. Hence seismic re-evaluation of various structures of the heavy water plant is carried out. Analysis of the heavy water plant structures was carried out for self weight, equipment load and earthquake load. Pressure loading was also considered in case of H 2 S storage tanks. Soil structure interaction effect was considered in the analysis. The combined stresses in the structures due to earthquake and dead load were checked with the allowable stresses. (author)

  17. Heavy water: a distinctive and essential component of CANDU

    International Nuclear Information System (INIS)

    Miller, A.I.; van Alstyne, H.M.

    1994-06-01

    The exceptional properties of heavy water as a neutron moderator provide one of the distinctive features of CANDU reactors. Although most of the chemical and physical properties of deuterium and protium (mass 1 hydrogen) are appreciably different, the low terrestrial abundance of deuterium makes the separation of heavy water a relatively costly process, and so of considerable importance to the CANDU system. World heavy-water supplies are currently provided by the Girdler-Sulphide process or processes based on ammonia-hydrogen exchange. Due to cost and hazard considerations, new processes will be required for the production of heavy water in and beyond the next decade. Through AECL's development and refinement of wetproofed catalysts for the exchange of hydrogen isotopes between water and hydrogen, a family of new processes is expected to be deployed. Two monothermal processes, CECE (Combined Electrolysis and Catalytic Exchange, using water-to-hydrogen conversion by electrolysis) and CIRCE (Combined Industrially Reformed hydrogen and Catalytic Exchange, based on steam reforming of hydrocarbons), are furthest advanced. Besides its use for heavy-water production, the CECE process is a highly effective technology for heavy-water upgrading and for tritium separation from heavy (or light) water. (author). 10 refs., 1 tab., 7 figs

  18. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  19. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

  20. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  1. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  2. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  3. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

    This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

  4. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  5. Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

    Science.gov (United States)

    Permana, Sidik

    2017-07-01

    A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

  6. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with ~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency and increase in

  7. Radiogenic Lead with Dominant Content of 208Pb: New Coolant and Neutron Moderator for Innovative Nuclear Facilities

    Directory of Open Access Journals (Sweden)

    A. N. Shmelev

    2011-01-01

    Full Text Available As a rule materials of small atomic weight (light and heavy water, graphite, and so on are used as neutron moderators and reflectors. A new very heavy atomic weight moderator is proposed—radiogenic lead consisting mainly of isotope 208Pb. It is characterized by extremely low neutron radiative capture cross-section (0.23 mbarn for thermal neutrons, i.e., less than that for graphite and deuterium and highest albedo of thermal neutrons. It is evaluated that the use of radiogenic lead makes it possible to slow down the chain fission reaction on prompt neutrons in a fast reactor. This can increase safety of the fast reactors and reduce as well requirements pertaining to the fuel fabrication technology. Radiogenic lead with high 208Pb content as a liquid-metal coolant of fast reactors helps to achieve a favorable (negative reactivity coefficient on coolant temperature. It is noteworthy that radiogenic lead with high 208Pb content may be extracted from thorium (as well as thorium-uranium ores without isotope separation. This has been confirmed experimentally by the investigations performed at San Paulo University, Brazil.

  8. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  9. Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

    International Nuclear Information System (INIS)

    Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

    2016-01-01

    KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

  10. Process for the extraction of tritium from heavy water

    International Nuclear Information System (INIS)

    Dombra, A.H.

    1984-01-01

    The object of the invention is achieved by a process for the extraction of tritium from a liquid heavy water stream comprising: contacting the heavy water with a countercurrent gaseous deuterium stream in a column packed with a water-repellent catalyst such that tritium is transferred by isotopic exchange from the liquid heavy water stream to the gaseous deuterium stream

  11. Fuel-Coolant Interactions: Visualization and Mixing Measurements

    International Nuclear Information System (INIS)

    Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

    2002-01-01

    Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

  12. Experience with dilute chemical decontamination in Indian Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Subramanian, Veena; Mittal, V.K.; Narasimhan, S.V.

    2010-01-01

    Dilute Chemical Decontamination (DCD) process has been used in several full system and components of nuclear coolant systems to effectively remove the radioactive contaminants that causes radiation field and consequent MANREM problem. The DCD process uses chemicals in very low concentrations (millimolar) and dissolves the oxide film along with the activity incorporated in the oxide film. In DCD process operated under the regenerative mode, the chemical formulation spent in the process of oxide dissolution is replenished by passing through cation exchange columns. Finally, after achieving sufficient decontamination of the system/component, the added decontamination chemicals along with the activities and metal ions released during the process are removed by mixed bed ion exchange columns and the system is restored to normal operating condition in few days time. In PHWRs, the regenerative DCD process is applied for full primary coolant system decontamination. The chemicals are added directly to the heavy water coolant with the fuel in the core. In Indian PHWRs (MAPS-1 and 2, RAPS-1 and 2, NAPS-1 and 2 and KAPS-1), the process has been applied eleven times. A chemical formulation based on NTA, Citric acid and Ascorbic acid has been applied seven times with good results. Decontamination factors in the range 2-30 have been obtained in different components with good MANREM savings in the subsequent maintenance works. Efforts are on to modify the process to take care of the challenges posed by antimony isotope. An inhibitor (Rodine-92B) based process was successfully tested in NAPS-2 for removing antimony isotopes ( 122 Sb and 124 Sb). Further refining of the antimony removal process is being worked out. Similarly, the process is being modified to effectively remove the hotspot causing stellite particles in the moderator system of PHWRs. A permanganate based process has been developed and tested in several adjustor rod drive mechanisms in KAPS and NAPS. The experience of

  13. Heavy water pumps; Pumpe D{sub 2}O

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V; Nikolic, M

    1963-12-15

    Continuous increase of radiation intensity was observed on all the elements in the heavy water system during first three years of RA reactor operation. The analysis of heavy water has shown the existence of radioactive cobalt. It was found that cobalt comes from stellite, cobalt based alloy which was used for coating of the heavy water pump discs in order to increase resistance to wearing. Cobalt was removed from the surfaces due to friction, and transferred by heavy water into the reactor where it has been irradiated for 29 876 MWh up to 8-15 Ci/g. Radioactive cobalt contaminated all the surfaces of aluminium and stainless steel parts. This report includes detailed description of heavy water pumps repair, exchange of stellite coated parts, decontamination of the heavy water system, distillation of heavy water. [Serbo-Croat] U toku prve tri godine eksploatacije reaktora RA uocen je neprekidni porast intenziteta zracenja na svim elementima u teskovodnom sistemu. Analizom teske vode utvrdjeno je postojanje radioaktivnog kobalta. Ustanovljeno je da kobalt potice od stelita, legure na bazi kobalta kojim su presvuceni rukavci vratila teskovodnih pumpi radi otpornosi na habanje. Kobalt je trenjem skidan sa povrsina, u toku rada prenosen je teskom vodom u reaktor i ozracivan u toku 29 876 MWh do specificne aktivnosti 8-15 Ci/g. Radioaktivni kobalt je kontaminirao sve povrsine od aluminijuma i nerdjajuceg celika. Ovaj izvestaj sadrzi detaljan opis remonta pumpi, zamene delova teskovodnih pumpi novim delovima bez stelitnog sloja, dekontaminacije teskovodnog sistema, destilacije teske vode.

  14. A Drinking Water Sensor for Lead and Other Heavy Metals.

    Science.gov (United States)

    Lin, Wen-Chi; Li, Zhongrui; Burns, Mark A

    2017-09-05

    Leakage of lead and other heavy metals into drinking water is a significant health risk and one that is not easily detected. We have developed simple sensors containing only platinum electrodes for the detection of heavy metal contamination in drinking water. The two-electrode sensor can identify the existence of a variety of heavy metals in drinking water, and the four-electrode sensor can distinguish lead from other heavy metals in solution. No false-positive response is generated when the sensors are placed in simulated and actual tap water contaminated by heavy metals. Lead detection on the four-electrode sensor is not affected by the presence of common ions in tap water. Experimental results suggest the sensors can be embedded in water service lines for long-time use until lead or other heavy metals are detected. With its low cost (∼$0.10/sensor) and the possibility of long-term operation, the sensors are ideal for heavy metal detection of drinking water.

  15. Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

    Science.gov (United States)

    Salamon, V.; Senthil kumar, D.; Thirumalini, S.

    2017-08-01

    The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

  16. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    The last decade worldwide experience of PWRs and WWERs confirms the trends for the improvement of the nuclear power industry electricity production through the implementation of high burn-up or high fuel duty, which are usually accompanied with the usage of UO 2 fuel with higher content of 235 U - 4.0% - 4.5% (5.0%). It was concluded that the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of the core are the primary and basic factors accompanying the implementation of uranium fuel with higher 235 U content, aiming extended fuel cycles and higher burn-up of the fuel in Pressurized Water Reactors. As main consequences of the presence of these factors the modifications of chemical / electrochemical environments of nuclear fuel cladding- and reactor coolant system- surfaces are evaluated. These conclusions are the reason for: 1) The determination of the choices of the type of fuel cladding materials in respect with their enough corrosion resistance to the specific fuel cladding environment, created by the presence of SNB; 2) The development and implementation of primary circuit water chemistry guidelines ensuring the necessary low corrosion rates of primary circuit materials and limitation of cladding deposition and out-of-core radioactivity buildup; 3) Implementation of additional neutron absorbers which allow enough decrease of the initial concentration of H 3 BO 3 in coolant, so that its neutralization will be possible with the permitted alkalising agent concentrations. In this paper the specific features of WWER-1000 units in Bulgarian Nuclear Power Plant; use of 235 U higher enriched fuel in the WWER-1000 reactors in the Kozloduy NPP; coolant water chemistry and radiochemistry plant data during the power operation period of the Kozloduy NPP Unit 5, 15 th fuel cycle; evaluation of the approaches and results by the conversion of the WWER-1000 Units at the Kozloduy NPP to the uranium fuel with 4.3% 235 U as

  17. Alkali metal and ammonium chlorides in water and heavy water (binary systems)

    CERN Document Server

    Cohen-Adad, R

    1991-01-01

    This volume surveys the data available in the literature for solid-fluid solubility equilibria plus selected solid-liquid-vapour equilibria, for binary systems containing alkali and ammonium chlorides in water or heavy water. Solubilities covered are lithium chloride, sodium chloride, potassium chloride, rubidium chloride, caesium chloride and ammonium chloride in water and heavy water.

  18. 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

  19. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    A. Rama Rao

    2008-04-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  20. Phytoremediation of water bodies contaminated with radioactive heavy metal

    International Nuclear Information System (INIS)

    Yan Zhen; Yuan Shichao; Ling Hui; Xie Shuibo

    2012-01-01

    The sources of the radioactive heavy metal in the water bodies were analyzed. The factors that affect phyto remediation of water contaminated with radioactive heavy metal were discussed. The plant species, mechanism and major technology of phyto remediation of water contaminated with radioactive heavy metal were particularly introduced. The prospective study was remarked. (authors)

  1. Fuel and heavy water availability

    International Nuclear Information System (INIS)

    1980-01-01

    The general guidelines for the Working Group's evaluation of the availability of nuclear fuel and heavy water were set at the Organizing Conference of the International Nuclear Fuel Cycle Evaluation (INFCE), which was held in Washington, United States of America, 19-21 October 1977. The agreed technical and economic scope for the evaluation was to: (1) Estimate needs for nuclear energy and correlated needs for uranium and heavy water according to different fuel cycle strategies; (2) Review uranium availability with specific regard to: Assessment of resources and production capacities; policies and incentives for encouraging exploration and production including joint ventures; marketing policies and/or guarantees of sales for companies investing in exploration and production; marketing policies and/or guarantees of supply for utilities; technical development of exploration, mining and milling methods; (3) Review heavy water availability; (4) Review thorium availability; (5) Consider special needs of developing countries. The illustrations of availability and requirements developed in this report do provide a useful framework for considering future options and alternatives for the development of nuclear power

  2. Production of heavy water by photodesorption

    International Nuclear Information System (INIS)

    Gangwer, T.; Goldstein, M.K.

    1976-01-01

    Research has recently brought attention to the laser as a tool for isotope enrichment. So far the main thrust of this effort has been toward uranium enrichment; however, numerous successes in other areas have been demonstrated. Isotopes of boron, sulfur, chlorine, and carbon have been separated. A new technique is proposed for laser isotope enrichment. The technique, referred to as photodesorption, involves selective isotopic excitation of molecules adsorbed on a surface such that an enrichment results from subsequent physical or chemical events undergone by the excited molecules. The specific processes of concern are the physical photodesorption enrichment of heavy water from light water and tritiated water from heavy water. The ability to work directly with water molecules has significant advantages for a commercial process. A photodesorption enrichment process has been forumulated and some analyses have been performed. This process is described and some preliminary cost estimates are made which assume successful accomplishment of the major R and D objectives of the new process. The results indicate that the process has the promise of a significant reduction in the cost of heavy water and that further study is warranted

  3. Heat transfer and fluid flow aspects of fuel--coolant interactions

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1978-09-01

    A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

  4. The heavy water production plant at Arroyito, Argentina

    International Nuclear Information System (INIS)

    Ecabert, R.

    1984-01-01

    The author describes the construction of an industrial heavy water production plant (Planta Industrial de Agua Pesada, PIAP) in Argentina. The heavy water enrichment is based on a hydrogen/ammonia isotope exchange. (Auth.)

  5. Standards for heavy water concentration determinations in light water

    International Nuclear Information System (INIS)

    Varlam, M.; Steflea, D.; Pavelescu, M.

    1995-01-01

    The paper presents a method to prepare heavy water -light water standards within the range 144 ppm - 1%. A formula for computing standards concentration based on initial concentration of D 2 O and distilled water is given

  6. Analysis of radiation exposure during creep adjustment to the coolant channels at Madras Atomic Power Station

    International Nuclear Information System (INIS)

    Varadhan, R.S.; Venkataramana, K.; Kannan, R.K.; Sreekumaran Nair, B.; Chudalayandi, K.

    1994-01-01

    In pressurised heavy water reactors the coolant channels made of zircaloy-2 undergo creep deformation used intense neutron irradiation in the reactor core. In order to measure and provide for the changes in the dimensions, base line data of internal diameters, sag and length of the 306 coolant channels are measured as pre service inspection (PSI) before the reactor is loaded with fuel prior to criticality. Subsequently as part of in service inspection (ISI), axial creep of every channel is measured in every annual shutdown of the reactor and creep adjustment is done on those channels where creep expansion margin for the next one year operation is low. A study was carried out to assess the radiological impact of the job at Madras Atomic Power Station (MAPS). Various measures adopted for reducing the individual and collective doses on the job are discussed in this report. (author). 3 refs., 2 tabs

  7. Overcoming technology - obsolescence: a case study in Heavy Water Plant

    International Nuclear Information System (INIS)

    Gupta, O.P.; Sonde, R.R.; Wechalekar, A.K.

    2002-01-01

    Ammonia based Heavy Water Plants in India are set up essentially in conjunction with fertiliser plants for the supply of feed synthesis gas. Earlier ammonia was being produced in fertiliser plants using high-pressure technology which was highly energy intensive. However with fast developments in the field of production of ammonia, fertiliser plants are switching over to low pressure technology. Ammonia based heavy water plants have to operate on pressures corresponding to that of fertiliser plants. Due to low pressures in production of ammonia, heavy water plants would also be required to operate at low pressures than the existing operating pressures. This problem was faced at Heavy Water Plant at Baroda where GSFC supplying synthesis gas switched over to low pressure technology making it imperative on the part of Heavy Water Board to carry out modification to the main plant for continued operation of Heavy Water Plant, Baroda. Anticipating similar problems due to production of ammonia at lower pressures in other fertiliser plants linked to existing Heavy Water Plants, it became necessary for HWB to develop water ammonia front end. The feed in such a case would be water instead of synthesis gas. This would enable HWB to dispense with dependence on fertiliser plants especially if grass-root ammonia based heavy water plants are to be set up. Incorporation of water ammonia front end would enable HWB to de link ammonia based heavy water plants with fertiliser plants. This paper discusses the advantage of de linking heavy water plant respective fertiliser plant by incorporating water ammonia front end and technical issues related to front end technology. A novel concept of ammonia absorption refrigeration (AAR) was considered for the process integration with the front end. The incorporation of AAR with water ammonia front-end configuration utilizes liquid ammonia refrigerant to generate refrigeration without additional energy input which otherwise would have been

  8. Electrolytic process for upgrading heavy water (Preprint No. PD-16)

    International Nuclear Information System (INIS)

    Rammohan, K.; Sadhukhan, H.K.

    1989-04-01

    In the reactor system the heavy water gets depleted in concentration due to leakages, intermixing and vapour collection in boiler vault system etc. Electrolysis of water was used as a secondary plant to enrich the dilute heavy water produced in the primery plant by hydrogen-sulfide-water exchange process. The studies made in the development of this process for the upgrading of downgra ded heavy water by setting up a full size Electrolyser Test Assembly are discussed a nd complete design of a heavy water upgrading plant based on electrolytic process for MAPS and NAPP is described. (author). 7 refs., 5 figs

  9. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  10. Primary Coolant pH Control for Soluble Boron-Free PWRs

    International Nuclear Information System (INIS)

    Cheon, Yang Ho; Lee, Nam Yeong; Park, Byeong Ho; Park, Seong Chan; Kim, Eun Kee

    2015-01-01

    These should be considered when evaluating and designing the operating pH program for nuclear power plants. This paper discusses the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water system for soluble boron pressurized water reactor (PWR) plants. Finally, the objective of this work is to study primary coolant pH control for soluble boron-free PWR plants. This paper reviewed the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water chemistry system for soluble boron PWR plants. The new chemistry trend for the primary coolant is towards adaption of the constant and elevated chemistry. Finally, this work studied primary coolant pH control for soluble boron-free PWR plants. The ammonia-based water chemistry related to pH control for boron-free PWR plants was discussed. The ammonia-based water chemistry is not recommended to avoid fluctuation of the pH value by ammonia radiolysis and to reduce C-14 production in reactor coolant from reaction with dissolved nitrogen. Also, the potassium-based water chemistry related to pH control for boron-free PWR plants was discussed. KOH has a potential as an alternative pH control agent for soluble boron-free PWR plants. The potassium-based water chemistry related to pH control is recommended for boron-free operation as follows

  11. Refurbishment of the technical water system, Task 3.08/04-09; Podzadatak 3.08/04-09 Remont sistema tehnicke vode

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, M; Milic, J [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    This report deals with repair and maintenance of the secondary coolant loop (technical water system) of the RA reactor. The need of proper functionality of the coolant system demanded inspection of the centrifugal pumps on Danube, cleaning of the pools for removal of sludge from the technical water; cleaning of the main heat exchangers and heavy water heat exchangers and inspection of all the parts of the cooling system situated in the reactor building.

  12. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  13. Alternate applications of heavy water in biological and technological fields

    International Nuclear Information System (INIS)

    Bhaskaran, M.; Prakash, R.

    2005-01-01

    Deuterium and its various compounds like heavy water exhibit distinctly different properties when compared to hydrogen and its compounds. The differences in properties are due to the primary and secondary isotopic effects. Though heavy water has been used solely for nuclear applications so far, its applications in life sciences and high technology areas are fast emerging. Heavy Water Board has taken up development of alternate applications of heavy water. The study taken up has indicated superior thermal stability for oral polio vaccine prepared in heavy water. This study has revealed various opportunities for application of heavy water or deuterium in life sciences and the paper dwells on these possibilities. The higher stability of compounds with deuterium has also brought in its applications in various high technology areas. These are mainly in micro electronics. Use of deuterium in manufacture of high quality optical fibres has already been established. These are also included in the paper. (author)

  14. Characterization and treatment options for high TOC heavy water

    International Nuclear Information System (INIS)

    Evans, D.; Leilabadi, A.; Rudolph, A.; Williams, D.

    2007-01-01

    High total organic carbon (TOC) and high conductivity contamination in heavy water feed present serious problems for the operation of heavy water upgrader facilities. The authors describe the chemical analysis of a particular batch of contaminated heavy water which had resisted standard clean-up procedures. After chemical characterization, a special clean-up plan was developed and successfully tested in the laboratory, followed by its implementation at site. (author)

  15. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  16. Iron crud supply device to reactor coolant

    International Nuclear Information System (INIS)

    Baba, Takao.

    1993-01-01

    In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

  17. Thermal conductivity coefficients of water and heavy water in the liquid state up to 3700C

    International Nuclear Information System (INIS)

    Le Neindre, B.; Bury, P.; Tufeu, R.; Vodar, B.

    1976-01-01

    The thermal conductivity coefficients of water and heavy water of 99.75 percent isotopic purity were measured using a coaxial cylinder apparatus, covering room temperature to their critical temperatures, and pressures from 1 to 500 bar for water, and from 1 to 1000 bar for heavy water. Following the behavior of the thermal conductivity coefficient of water, which shows a maximum close to 135 0 C, the thermal conductivity coefficient of heavy water exhibits a maximum near 95 0 C and near saturation pressures. This maximum is displaced to higher temperatures when the pressure is increased. Under the same temperature and pressure conditions the thermal conductivity coefficient of heavy water was lower than for water. The pressure effect was similar for water and heavy water. In the temperature range of our experiments, isotherms of thermal conductivity coefficients were almost linear functions of density

  18. Survival of tumor-bearing mice exposed to heavy water or heavy water plus methotrexate

    International Nuclear Information System (INIS)

    Laissue, J.A.; Buerki, H.; Berchtold, W.

    1982-01-01

    Moderate body deuteration combined with a cytostatic drug [methotrexate (MTX)] significantly increases the survival time of young adult DBA/2 mice bearing transplantable P815. L5178Y, or L1210 tumors. Neoplastic cells were grown in vitro from tumor stock and injected i.p. into mice from two groups, one drinking tap water, and other drinking 30% heavy water in tap water. One-half of the animals in each of these two groups was given a single injection of MTX (4 mg/kg body weight) on 3 consecutive days per week. At death, extension of primary and metastatic tumors was examined and was found to be macro- and microscopically comparable in the corresponding groups. The mean survival time of untreated mice drinking tap water was about 2 weeks following injection of the fast-growing P815, L5178Y, or L1210 (V) tumors and approximately 5 weeks after injection of cells from a slower-growing L1210 subline. Body deuteration alone roughly doubled the survival time solely of mice bearing this L1210 subline. Treatment with MTX approximately doubled the mean survival time of hosts bearing one of the fast-growing tumors. Combined treatment with heavy water and MTX increased the mean survival time of the mice in all groups by 15 to 125% as compared to control values. The reasons for this effect are unknown. However, heavy water has been shown to exert antimitotic activity and to depress the incorporation of radioactive precursors into DNA of proliferating mammalian cells. The depression of antibody formation following antigenic stimulation and the reduction in numbers of nonneoplastic lymphoid cells of mice following moderate body deuteration may have contributed to the enhancement of MTX activity in addition to other effects of deuterium

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

    International Nuclear Information System (INIS)

    1989-07-01

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

  20. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    Osmachkin, V.S.; Sokolov, I.N.

    1998-01-01

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  1. Design of a thorium fuelled Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    2009-01-01

    Full text: The main objective for development of Advanced Heavy Water Reactor (AHWR) is to demonstrate thorium fuel cycle technologies, along with several other advanced technologies required for next generation reactors, so that these are readily available in time for launching the third stage. The AHWR under design is a 300 MWe vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The fuel consists of (Th-Pu)O 2 and ( 233 ThU)O 2 pins. The fuel cluster is designed to generate maximum energy out of 233 U, which is bred in-situ from thorium and has a slightly negative void coefficient of reactivity, negative fuel temperature coefficient and negative power coefficient. For the AHWR, the well -proven pressure tube technology and online fuelling have been adopted. Core heat removal is by natural circulation of coolant during normal operation and shutdown conditions. Thus, it combines the advantages of light water reactors and PHWRs and removes the disadvantages of PHWRs. It has several passive safety systems for reactor normal operation, decay heat removal, emergency core cooling, confinement of radioactivity etc. The fuel cycle is based on the in-situ conversion of naturally available thorium into fissile 233 U in self sustaining mode. The uranium in the spent fuel will be reprocessed and recycled back into the reactor. The plutonium inventory will be kept a minimum and will come from fuel irradiated in Indian PHWRs. The 233 U required initially can come from the fast reactor programme or it can be produced by specially designing the initial core of AHWR using (Th,Pu)MOX fuel. There will be gradual transition from the initial core which will not contain any 233 U to an equilibrium core, which will have ( 233 U, Th) MOX fuel pins also in a composite cluster. The self sustenance is being achieved by a differential fuel loading of low and a relatively higher Pu in the composite clusters. The AHWR burns the

  2. assessment of heavy metals concentration in drinking water ...

    African Journals Online (AJOL)

    userpc

    guidelines (WHO 2005). Findings suggest that continues water quality monitoring should be carried out to check the concentration levels of heavy metals in that area, to prevent them from been above the limit of WHO. Keywords: Atomic Absorption Spectrophotometers, Heavy Metals, Water, Kauru Local. Government Area.

  3. Experimental investigation of heat transfer potential of Al2O3/Water-Mono Ethylene Glycol nanofluids as a car radiator coolant

    Directory of Open Access Journals (Sweden)

    Dattatraya G. Subhedar

    2018-03-01

    Full Text Available In this research, the heat transfer potential of Al2O3/Water-Mono Ethylene Glycol nanofluids is investigated experimentally as a coolant for car radiators. The base fluid was the mixture of water and mono ethylene glycol with 50:50 proportions by volume. The stable nanofluids obtained by ultra-sonication are used in all experiments. In this study nanoparticle volume fraction, coolant flow rate, inlet temperature used in the ranges of 0.2–0.8%, 4–9 l per minute and 65–85 °C. The results show that the heat transfer performance of radiator is enhanced by using nanofluids compared to conventional coolant. Nanofluid with lowest 0.2% volume fraction 30% rise in heat transfer is observed. Also the estimation of reduction in frontal area of radiator if base fluid is replaced by Nanofluid is done which will make lighter cooling system, produce less drag and save the fuel cost.

  4. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  5. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Nanis, R.

    2000-01-01

    Hydrogen atom has two isotopes: deuterium 1 H 2 and tritium 1 H 3 . The deuterium oxide D 2 O is called heavy water due to its density of 1105.2 Kg/m 3 . Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D 2 O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D 2 O management is required to preserve it. (author)

  6. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  7. Heavy water radiolysis and chemistry control of the Fugen Nuclear Power Station

    International Nuclear Information System (INIS)

    Ibuki, Y.; Kitabata, T.; Kato, T.

    1989-01-01

    A computer analysis for heavy water radiolysis clarified the mechanism of the heavy water radiolysis rate change with impurities in the heavy water and cover gas, helium. The mechanism is supported by over ten years' operational data of the heavy water radiolysis in the Fugen nuclear power station. (author)

  8. Method of eliminating cruds in the primary coolants of reactors

    International Nuclear Information System (INIS)

    Tamura, Takaaki.

    1984-01-01

    Purpose: To eliminate cruds in the primary coolants by using rind of onions or peanuts. Method: Since cruds contained in the reactor primary coolants increase the radioactive exposure to reactor operators, they have been intended to remove by ion exchange resins. In this invention, rind of onions or peanuts are crushed into an adequate particle size and packed into an absorption column instead of ion exchange resins into which primary coolants are circulated. The powderous onions or peanuts rind contain glucoside such as cosmosiin and has an effect of cationic exchanger, they satisfactorily catch heavy metals such as Fe and Cu. They have an excellent filtering effect even under a high pH condition and are excellent in economical point of view. They can be decrease the volume of the absorption column, reduce their devolume after use through corrosion and easily subjected to waste procession through oxidizing combustion in liquid. (Nakamoto, H.)

  9. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  10. Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

    2000-04-01

    The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

  11. 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.)

  12. Dynamic response of INTOR/NET blankets after coolant tube rupture

    International Nuclear Information System (INIS)

    Klippel, H.T.

    1985-01-01

    The dynamic response of different water-cooled liquid Li 17 Pb 83 breeder blanket modules has been calculated to study the potential of these modules in case of coolant tube rupture. Numerical calculations with the code PISCES have been carried out taking into account the fluid-structure interaction and the elasto-plastic behaviour of the structural material. The results show that for inert coolant characteristics the proposed conceptual designs for NET and INTOR have sufficient resistance against coolant tube rupture but when taking into account energy release due to chemical reaction of water with LiPb-alloy up to doubling of the wall thickness has to be envisaged to guarantee structural reliability. (orig.)

  13. One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  14. Determination of blade-to-coolant heat-transfer coefficients on a forced-convection, water-cooled, single-stage turbine

    Science.gov (United States)

    Freche, John C; Schum, Eugene F

    1951-01-01

    Blade-to-coolant convective heat-transfer coefficients were obtained on a forced-convection water-cooled single-stage turbine over a large laminar flow range and over a portion of the transition range between laminar and turbulent flow. The convective coefficients were correlated by the general relation for forced-convection heat transfer with laminar flow. Natural-convection heat transfer was negligible for this turbine over the Grashof number range investigated. Comparison of turbine data with stationary tube data for the laminar flow of heated liquids showed good agreement. Calculated average midspan blade temperatures using theoretical gas-to-blade coefficients and blade-to-coolant coefficients from stationary-tube data resulted in close agreement with experimental data.

  15. Recent developments in coolant systems for Indus Accelerator Complex at RRCAT, Indore

    International Nuclear Information System (INIS)

    Nanda, Dipankar; Tiwari, Bablu; Pandey, R.M.

    2015-01-01

    Scarcity of fresh water forces mankind to explore other possible water sources that can meet the increasing demand of coolants in industries, R and D sectors and other establishments where water is used as coolant. It also becomes a challenge for water chemist to control water chemistry to keep the equipments/devices intact during its operation using water as coolant. Deionised (DI) and soft water have been used as coolants for Indus Accelerator Complex, RRCAT, Indore. DI water is produced and its quality is maintained either by conventional ion exchange method or a hybrid method of membrane separation and ion exchange technique. This requires handling of corrosive chemicals, manpower, space for plant installation, and a long array of water treatment units. CSL has implemented the idea of rain water harvesting to produce DI water after systematic studies in laboratory. The concerning issues are reduced to almost one-fourth by using rain water to produce DI water. The harvesting system has been in use for last three years. Heat is dissipated into air by evaporation of soft water in cooling tower. Requirement of soft water makeup has been estimated to be about 40,000 ltrs. / day (max.) if the machine is operated at its designed specifications. Non-availability of soft water (which circulates in open loop) may lead to shut down like situation and looking for alternate source becomes quite essential. Laboratory studies (water analysis and treatment) on sewage water (available 1,00,000 ltrs/day) from RRCAT colony as a possible source of producing soft water show promising result. (author)

  16. Heavy water isotopic rectification in the ''ORPHEE'' reactor. SACLAY studies Centre

    International Nuclear Information System (INIS)

    Lejeune, P.; Breant, P.

    1993-01-01

    ORPHEE reactor supplies neutron beams, which are got back in a heavy water reflector. The neutron beams intensity depends on the reflector quality which is determined by the isotopic content of the heavy water. The deuterium submitted to core irradiation changes in radioactive tritium which must be eliminated largely for reasons of safety. The column must keep the heavy water isotopic content of the reflector to a value higher than 99.8% by eliminating light water by fractional distillation or rectification. This column is also used for the tritium elimination of heavy water. 13 figs

  17. Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

    International Nuclear Information System (INIS)

    Williams, C.L.; Beus, S.G.

    1980-05-01

    Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values

  18. Process for the preparation of ammonia and heavy water

    International Nuclear Information System (INIS)

    Mandrin, C.

    1980-01-01

    A process for the production of ammonia and heavy water comprises the steps of enriching a flow of water with deuterium in a monothermal isotropic process; supplying a first portion of the deuterium-enriched water to a heavy water preparation plant to produce heavy water and hydrogen; storing a second portion of the deuterium-enriched water substantially without interruption during the colder half of a year; electrolytically dissociating the stored deuterium-enriched water substantially without interruption during the wamer half of a year to form hydrogen; storing a portion of the electrolytically-produced hydrogen during said warmer half of a year while supplying the remainder to a synthesis circuit of a synthesizing plant and subsequently supplying the stored hydrogen to the synthesis circuit during said colder half of a year; removing some of the synthesis gas mixture from the synthesis circuit of the synthesizing plant; burning the removed synthesis gas mixture with air to produce a mixture consisting mainly of water and nitrogen; thereafter condensing and separating the water from the mixture of water and nitrogen; supplying the nitrogen of the mixture of water and nitrogen, the hydrogen from the heavy water preparation plant and the electrolytically-produced hydrogen to the synthesis circuit of the synthesizing plant to produce ammonia; and collecting deuterium-depleted water resulting from said burning step and feeding the collected deuterium-depleted water into the monothermal process

  19. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  20. Heavy water at Trail, British Columbia

    International Nuclear Information System (INIS)

    Arsenault, J.E.

    2006-01-01

    Today Canada stands on the threshold of a nuclear renaissance, based on the CANDU reactor family, which depends on heavy water as a moderator and for cooling. Canada has a long history with heavy water, with commercial interests beginning in 1934, a mere two years after its discovery. At one time Canada was the world's largest producer of heavy water. The Second World War stimulated interest in this rather rare substance, such that the worlds largest supply (185 kg) ended up in Canada in 1942 to support nuclear research work at the Montreal Laboratories of the National Research Council. A year later commercial production began at Trail, British Columbia, to support work that later became known as the P-9 project, associated with the Manhattan Project. The Trail plant produced heavy water from 1943 until 1956, when it was shut down. During the war years the project was so secret that Lesslie Thomson, Special Liaison Officer reporting on nuclear matters to C.D. Howe, Minister of Munitions and Supply, was discouraged from visiting Trail operations. Thomson never did visit the Trail facility during the war. In 2005 the remaining large, tall concrete exchange tower was demolished at a cost of about $2.4 million, about the same as it cost to construct the facility about 60 years ago. Thus no physical evidence remains of this historic facility and another important artifact from Canada's nuclear history has disappeared forever. It is planned to place a plaque at the site at some point in the future. (author)

  1. Heavy water at Trail, British Columbia

    Energy Technology Data Exchange (ETDEWEB)

    Arsenault, J.E. [Ontario (Canada)

    2006-09-15

    Today Canada stands on the threshold of a nuclear renaissance, based on the CANDU reactor family, which depends on heavy water as a moderator and for cooling. Canada has a long history with heavy water, with commercial interests beginning in 1934, a mere two years after its discovery. At one time Canada was the world's largest producer of heavy water. The Second World War stimulated interest in this rather rare substance, such that the worlds largest supply (185 kg) ended up in Canada in 1942 to support nuclear research work at the Montreal Laboratories of the National Research Council. A year later commercial production began at Trail, British Columbia, to support work that later became known as the P-9 project, associated with the Manhattan Project. The Trail plant produced heavy water from 1943 until 1956, when it was shut down. During the war years the project was so secret that Lesslie Thomson, Special Liaison Officer reporting on nuclear matters to C.D. Howe, Minister of Munitions and Supply, was discouraged from visiting Trail operations. Thomson never did visit the Trail facility during the war. In 2005 the remaining large, tall concrete exchange tower was demolished at a cost of about $2.4 million, about the same as it cost to construct the facility about 60 years ago. Thus no physical evidence remains of this historic facility and another important artifact from Canada's nuclear history has disappeared forever. It is planned to place a plaque at the site at some point in the future. (author)

  2. Chemistry in production of heavy water and industrial solvents

    International Nuclear Information System (INIS)

    Thomas, P.G.

    2015-01-01

    Industries are the temples of modern science built on the robust foundation of science and technology. The genesis of giant chemical industries is from small laboratories where the scientific thoughts are fused and transformed into innovative technologies Heavy water production is an energy intensive giant chemical industry where various hazardous and flammable chemicals are handled, extreme operating conditions are maintained and various complex chemical reactions are involved. Chemistry is the back bone to all chemical industrial activities and plays a lead role in heavy water production also. Heavy Water Board has now mastered the technology of design, construction, operation and maintenance of Heavy Water plants as well as fine tuning of the process make it more cost effective and environment friendly. Heavy Water Board has ventured into diversified activities intimately connected with our three stages of Nuclear Power Programme. Process development for the production of nuclear grade solvents for the front end and back end of our nuclear fuel cycle is one area where we have made significant contributions. Heavy Water Board has validated, modified and fine-tuned the synthesis routes for TBP, D2EHPA, TOPO, TAPO TIAP, DNPPA, D2EHPA-II, DHOA etc and these solvents were accepted by end users. Exclusive campaigns were carried out in laboratory scale, bench scale and pilot plant scale before scaling up to industrial scale. The process chemistry is understood very well and chemical parameters were monitored in every step of the synthesis. It is a continual improvement cycle where fine tuning is carried out for best quality and yield of product at lowest cost. In this presentation, an attempt is made to highlight the role of chemistry in the production of Heavy Water and industrial solvents

  3. Method and apparatus for enrichment or upgrading heavy water

    International Nuclear Information System (INIS)

    Butler, J.P.; Hammerli, M.

    1979-01-01

    A method and apparatus for upgrading and final enrichment of heavy water are described, comprising means for contacting partially enriched heavy water feed in a catalyst column with hydrogen gas (essentially D 2 ) originating in an electrolysis cell so as to enrich the feed water with deuterium extracted from the electrolytic hydrogen gas and means for passing the deuterium enriched water to the electrolysis cell. (author)

  4. The Canadian heavy water situation

    International Nuclear Information System (INIS)

    Dahlinger, A.

    The Canadian heavy water industry is analyzed. Supply and demand are predicted through 1985. Pricing is broken down into components. Backup R and D contributes greatly to process improvements. (E.C.B.)

  5. Tritium separation from heavy water using electrolysis

    International Nuclear Information System (INIS)

    Ogata, Y.; Sakuma, Y.; Ohtani, N.; Kodaka, M.

    2001-01-01

    A tritium separation from heavy water by the electrolysis using a solid polymer electrode (SPE) was specified on investigation. The heavy water (∼10 Bq g -1 ) and the light water (∼70 Bq g -1 ) were electrolysed using an electrolysis device (Tripure XZ001, Permelec Electrode Ltd.) with the SPE layer. The cathode was made of stainless steel (SUS314). The electrolysis was carried out at 20 A x 60 min, with the electrolysis temperature at 10, 20, or 30degC, and 15 A x 80 min at 5degC. The produced hydrogen and oxygen gases were recombined using a palladium catalyst (ND-101, N.E. Chemcat Ltd.) with nitrogen gas as a carrier. The activities of the water in the cell and of the recombined water were analyzed using a liquid scintillation counter. The electrolysis potential to keep the current 20 A was 2-3 V. The yields of the recombined water were more than 90%. The apparent separation factors (SF) for the heavy water and the light water were ∼2 and ∼12, respectively. The SF value was in agreement with the results in other work. The factors were changed with the cell temperature. The electrolysis using the SPE is applicable for the tritium separation, and is able to perform the small-scale apparatus at the room temperature. (author)

  6. Improved method of degassing of feed water at Heavy Water Plant, Kota

    International Nuclear Information System (INIS)

    Krishnan, G.K.; Agrawal, A.K.

    1994-01-01

    Heavy Water Plant (Kota) processes 450 MT/hr of feed water as the source of deuterium using water/hydrogen sulphide exchange process for the production of heavy water. Plant design has limited the ingress of dissolved oxygen in feed water to 0.2 ppm. However, even this low limit on dissolved oxygen has been found unacceptable during plant operation as over an operational period of 3-4 years accumulation of sulphur due to oxidation of hydrogen sulphide on exchange tower trays poses major operational problems. This paper discusses the results of nitrogen injection used for reducing the ingress of dissolved oxygen in the feed water system of the plant. (author)

  7. Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

    The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

  8. High resolution conductometry for isotopic assay of deuterium in mixtures of heavy water and light water

    International Nuclear Information System (INIS)

    Ananthanarayanan, R.; Sahoo, P.; Murali, N.

    2014-01-01

    A PC based high resolution conductivity monitoring technique has been deployed for determination of isotopic purity of heavy water in samples containing heavy water and light water mixtures using pulsating sensor based conductivity monitoring instrument. The technique involves accurate determination of conductivities of a series of specially treated heavy water and light water mixtures of various compositions at a constant solution temperature. The shift in conductivity (Δκ), which is the difference between conductivities of composite mixture after and before the formation of a typical complex compound (boric acid–mannitol complex in this case), shows a smooth and reproducible decreasing trend with increase in percentage composition of heavy water. This relation, which is obtained by appropriate calibration, is used in the software program for direct display of isotopic purity of heavy water. The technique is examined for determination of percentage composition of heavy water in the entire range of concentration (0-100 %) with reasonable precision (relative standard deviation, RSD ≤1.5 %). About 1 mL of sample is required for each analysis and analysis is completed within a couple of minutes after pretreatment of sample. The accuracy in measurement is ≤1.75 %. (author)

  9. Coolant controls of a PEM fuel cell system

    Science.gov (United States)

    Ahn, Jong-Woo; Choe, Song-Yul

    When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  11. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  12. Canadian heavy water production - 1970 to 1980

    International Nuclear Information System (INIS)

    Galley, M.R.

    1981-01-01

    In the last decade, heavy water production in Canada has progressed from the commissioning of a single unit plant in Nova Scotia to a major production industry employing 2200 persons and operating three plants with an aggregate annual production capability in excess of 1800 Mg. The decade opened with an impending crisis in the supply of heavy water due to failure of the first Glace Bay Heavy Water Plant and difficulty in commissioning the second Canadian plant at Port Hawkesbury. Lessons learned at this latter plant were applied to the Bruce plant where the first two units were under construction. When the Bruce units were commissioned in 1973 the rate of approach to design production rates was much improved, renewing confidence in Canada's ability to succeed in large scale heavy water production. In the early 1970's a decision was made to rehabilitate the Glace Bay plant using a novel flowsheet and this rebuilt plant commenced production in 1976. The middle of the decade was marked by two main events: changes in ownership of the operating plants and initiation of a massive construction program to support the forecast of a rapidly expanding CANDU power station construction program. New production units embodying the best features of their predecessors were committed at Bruce by Ontario Hydro and at La Prade, Quebec, by AECL. The high growth rate in electrical demand did not continue and some new plant construction was curtailed. The present installed production capacity will now probably be adequate to meet anticipated demand for the next decade. Canadian plants have now produced more than 7800 Mg of heavy water at a commercially acceptable cost and with a high degree of safety and compliance with appropriate environmental regulations

  13. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  14. 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

  15. Cell growth and protein synthesis of unicellular green alga Chlamydomonas in heavy water

    International Nuclear Information System (INIS)

    Ishida, M.R.

    1983-01-01

    The effects of heavy water on the cell growth and protein synthesis of the photoautotrophically growing Chlamydomonas cells were studied. The growth rate of the cells is inversely proportional to the concentrations of heavy water. The cells can barely live in 90% heavy water, but they die in 99.85% heavy water within a few days. Incorporation of 14 Cleucine into cells is markedly stimulated by heavy water of various concentrations between 30 and 99.85% in the case of the short time incubation. Contrary to this, in the long time incubation as several days, heavy water inhibits the protein synthesis. Such two modes of the protein synthetic activities are dependent upon the incubation time of the cells grown photoautotrophically in the heavy water media. (author)

  16. Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

    International Nuclear Information System (INIS)

    Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

    1976-01-01

    In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

  17. The Steam Generating Heavy Water Reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    An account is given of the SGHWR, the prototype of which was built by the United Kingdom Atomic Energy Authority at Winfrith, under the following headings: Introduction; origin of the SGHWR concept; conceptual design (choice of reactor type, steam cycle, reactor coolant system, nuclear behaviour, fuel design, core design, and protective, auxiliary and containment systems); operation and control (integrity of core cooling, reactivity control, power trimming, long term reactivity control, xenon override, load following, power shaping, spatial stability control, void coefficient); protective systems (breached coolant circuit trip, intact coolant circuits trip, power set-back trip); dynamic characteristics; reactor control; station control (decoupled control system, coupled control system, rate of response); Winfrith prototype (design and safety philosophy, conceptual features and parameters, reactor coolant system, protective systems, emergency core cooling, core structure, fuel design, vented containment). (U.K.)

  18. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  19. Electromagnetic radiation during electrolysis of heavy water

    International Nuclear Information System (INIS)

    Koval'chuk, E.P.; Yanchuk, O.M.; Reshetnyak, O.V.

    1994-01-01

    The radiation in the visible and ultraviolet spectral regions during electrolysis of heavy water on nickel and palladium cathodes was determined for the first time. A sharp jump of the intensity photon flow was observed at a current density of higher than 125 mA/cm 2 . A hypothesis about the relation of the electrochemiluminescence phenomenon during electrolysis of heavy water with the formation of fresh surfaces in consequence of the hydrogenous corrosion of the cathode material is formulated. ((orig.))

  20. Heavy metals in drinking water: Occurrences, implications, and future needs in developing countries

    International Nuclear Information System (INIS)

    Chowdhury, Shakhawat; Mazumder, M.A. Jafar; Al-Attas, Omar; Husain, Tahir

    2016-01-01

    Heavy metals in drinking water pose a threat to human health. Populations are exposed to heavy metals primarily through water consumption, but few heavy metals can bioaccumulate in the human body (e.g., in lipids and the gastrointestinal system) and may induce cancer and other risks. To date, few thousand publications have reported various aspects of heavy metals in drinking water, including the types and quantities of metals in drinking water, their sources, factors affecting their concentrations at exposure points, human exposure, potential risks, and their removal from drinking water. Many developing countries are faced with the challenge of reducing human exposure to heavy metals, mainly due to their limited economic capacities to use advanced technologies for heavy metal removal. This paper aims to review the state of research on heavy metals in drinking water in developing countries; understand their types and variability, sources, exposure, possible health effects, and removal; and analyze the factors contributing to heavy metals in drinking water. This study identifies the current challenges in developing countries, and future research needs to reduce the levels of heavy metals in drinking water. - Highlights: • Co-exposure to multiple heavy metals in drinking water needs better understanding • Low-cost technologies for arsenic removal needs urgent attention • Protonated alginate needs further research for drinking water applications • Community level and PoU devices need improvement and cost reduction • Developing countries are most affected by heavy metals in drinking water

  1. Heavy metals in drinking water: Occurrences, implications, and future needs in developing countries

    Energy Technology Data Exchange (ETDEWEB)

    Chowdhury, Shakhawat, E-mail: Schowdhury@kfupm.edu.sa [Department of Civil and Environmental Engineering, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Mazumder, M.A. Jafar [Department of Chemistry, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Al-Attas, Omar [Department of Civil and Environmental Engineering, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Husain, Tahir [Faculty of Engineering and Applied Science, Memorial University of Newfoundland, St. John’s, NL (Canada)

    2016-11-01

    Heavy metals in drinking water pose a threat to human health. Populations are exposed to heavy metals primarily through water consumption, but few heavy metals can bioaccumulate in the human body (e.g., in lipids and the gastrointestinal system) and may induce cancer and other risks. To date, few thousand publications have reported various aspects of heavy metals in drinking water, including the types and quantities of metals in drinking water, their sources, factors affecting their concentrations at exposure points, human exposure, potential risks, and their removal from drinking water. Many developing countries are faced with the challenge of reducing human exposure to heavy metals, mainly due to their limited economic capacities to use advanced technologies for heavy metal removal. This paper aims to review the state of research on heavy metals in drinking water in developing countries; understand their types and variability, sources, exposure, possible health effects, and removal; and analyze the factors contributing to heavy metals in drinking water. This study identifies the current challenges in developing countries, and future research needs to reduce the levels of heavy metals in drinking water. - Highlights: • Co-exposure to multiple heavy metals in drinking water needs better understanding • Low-cost technologies for arsenic removal needs urgent attention • Protonated alginate needs further research for drinking water applications • Community level and PoU devices need improvement and cost reduction • Developing countries are most affected by heavy metals in drinking water.

  2. Improved method of degassing of feed water at Heavy Water Plant, Kota

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, G K; Agrawal, A K [Heavy Water Plant, Kota (India)

    1994-06-01

    Heavy Water Plant (Kota) processes 450 MT/hr of feed water as the source of deuterium using water/hydrogen sulphide exchange process for the production of heavy water. Plant design has limited the ingress of dissolved oxygen in feed water to 0.2 ppm. However, even this low limit on dissolved oxygen has been found unacceptable during plant operation as over an operational period of 3-4 years accumulation of sulphur due to oxidation of hydrogen sulphide on exchange tower trays poses major operational problems. This paper discusses the results of nitrogen injection used for reducing the ingress of dissolved oxygen in the feed water system of the plant. (author). 1 fig.

  3. Safety system in a heavy water detritiation plant

    International Nuclear Information System (INIS)

    Balteanu, O.; Stefan, I.; Retevoi, C.

    2003-01-01

    In a CANDU 6 type reactor a quantity of 55·10 15 Bq/year of tritium is generated, 95% being in the D 2 O moderator which can achieve a radioactivity of 2.5-3.5·10 12 Bq/kg. Tritium in heavy water contributes with 30-50% to the doses received by operation personnel and up to 20% to the radioactivity released in the environment. The large quantity of heavy water used in this type of reactors (500 tones) make storage very difficult, especially for environment. The extraction of tritium from tritiated heavy water of CANDU reactors solve the following problems: the radiation level in the operation area, the costs of maintenance and repair reduction due to reduction of personnel protection measures, the increase of NPP utilisation factor by shutdown time reduction for maintenance and repair, use the extracted tritium for fusion reactors and not for the last, lower costs and risk for storage heavy water waste. Heavy water detritiation methods, which currently are used in the industrial or experimental plant, are based on catalytic isotope exchange or electrolysis followed cryogenic distillation or permeation. The technology developed at Institute of Cryogenics and Isotope Separation is based upon catalytic exchange between tritiated water and deuterium, followed by cryogenic distillation of hydrogen isotopes. The nature of the fluids that are processed in detritiation requires the operation of the plant in safety conditions. The paper presents the safety system solution chose in order to solve this task, as well as a simulation of an incident and safety system response. The application software is using LabView platform that is specialised on control and factory automation applications. (author)

  4. Modification of water treatment plant at Heavy Water Plant (Kota)

    International Nuclear Information System (INIS)

    Gajpati, C.R.; Shrivastava, C.S.; Shrivastava, D.C.; Shrivastava, J.; Vithal, G.K.; Bhowmick, A.

    2008-01-01

    Heavy Water Production by GS process viz. H 2 S - H 2 O bi-thermal exchange process requires a huge quantity of demineralized (DM) water as a source of deuterium. Since the deuterium recovery of GS process is only 18-19%, the water treatment plant (WTP) was designed and commissioned at Heavy Water Plant (Kota) to produce demineralized water at the rate of 680 m 3 /hr. The WTP was commissioned in 1980 and till 2005; the plant was producing DM water of required quality. It was having three streams of strong cation resin, atmospheric degasser and strong anion exchange resin with co-current regeneration. In 2001 a new concept of layered bed resin was developed and engineered for water treatment plant. The concept was attractive in terms of saving of chemicals and thus preservation of environment. Being an ISO 9000 and ISO 14000 plant, the modification of WTP was executed in 2005 during major turn around. After modification, a substantial amount of acid and alkali is saved

  5. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Armstrong, J.; Hamilton, H.; Hyland, B. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

    2013-07-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  6. Lattice cell and full core physics of internally cooled annular fuel in heavy water moderated reactors

    International Nuclear Information System (INIS)

    Armstrong, J.; Hamilton, H.; Hyland, B.

    2013-01-01

    A program is underway at Atomic Energy of Canada Limited (AECL) to develop a new fuel bundle concept to enable greater burnups for PT-HWR (pressure tube heavy water reactor) cores. One option that AECL is investigating is an internally cooled annular fuel (ICAF) element concept. ICAF contains annular cylindrical pellets with cladding on the inner and outer diameters. Coolant flows along the outside of the element and through the centre. With such a concept, the maximum fuel temperature as a function of linear element rating is significantly reduced compared to conventional, solid-rod type fuel. The preliminary ICAF bundle concept considered in this study contains 24 half-metre long internally cooled annular fuel elements and one non-fuelled centre pin. The introduction of the non-fuelled centre pin reduces the coolant void reactivity (CVR), which is the increase in reactivity that occurs on voiding the coolant in accident scenarios. Lattice cell and full core physics calculations of the preliminary ICAF fuel bundle concept have been performed for medium burnups of approximately 18 GWd/tU using WIMS-AECL and reactor fuel simulation program (RFSP). The results will be used to assist in concept configuration optimization. The effects of radial and axial core power distributions, linear element power ratings, refuelling rates and operational power ramps have been analyzed. The results suggest that burnups of greater than 18 GWd/tU can be achieved in current reactor designs. At approximately 18 GWd/tU, expected maximum linear element ratings in a PT-HWR with online-refuelling are approximately 90 kW/m. These conditions would be prohibitive for solid-rod fuel, but may be possible in ICAF fuel given the reduced maximum fuel temperature as a function of linear element rating. (authors)

  7. A review of the UKAEA interest in heavy water reactors

    International Nuclear Information System (INIS)

    Symes, R.J.

    1983-01-01

    The chapter commences with a brief account of the history of heavy water production and then begins the story of the British use of this moderator in power reactors. This is equated with the introduction and development of the tube reactor as a distinct and important form of reactor construction in contrast with the perhaps better known vessel design that has tended to dominate reactor engineering to date. The account thus includes a succession of reactor designs including the gas and steam cooled heavy water systems in addition to the steam-generating heavy water reactor. The SGHWR was demonstrated by the construction of a substantial prototype, which continues in operation as a flexible and reliable electricity-generating plant. It was also, for a time, identified as the system to be used for Britain's third reactor programme. Today the successful Canadian CANDU power reactors represent the only penetration of heavy water reactor technology into large scale electricity generation. The range of research and experimental reactors using heavy water in their cores is reviewed. (author)

  8. Coolant material effect on the heat transfer rates of the molten metal pool with solidification

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1998-01-01

    Experimental studies on heat transfer and solidification of the molten metal pool with overlying coolant with boiling were performed. The simulant molten pool material is tin (Sn) with the melting temperature of 232 degree C. Demineralized water and R113 are used as the working coolant. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The Nusselt number and the Rayleigh number in the molten metal pool region of this study are compared between the water coolant case and the R113 coolant case. The experimental results for the water coolant are higher than those for R113. Also, the empirical relationship of the Nusselt number and the Rayleigh number is compared with the literature correlations measured from mercury. The present experimental results are higher than the literature correlations. It is believed that this discrepancy is caused by the effect of the heat loss to the environment on the natural convection heat transfer in the molten pool

  9. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

  10. Direction of Heavy Water Projects

    International Nuclear Information System (INIS)

    1984-07-01

    Summary of the activities performed by the Heavy Water Projects Direction of the Argentine Atomic Energy Commission from 1950 to 1983. It covers: historical data; industrial plant (based on ammonia-hydrogen isotopic exchange); experimental plant (utilizing hydrogen sulfides-water process); Module-80 plant (2-3 tons per year experimental plant with national technology) and other related tasks on research and development (E.A.C.) [es

  11. Blade-to-coolant heat-transfer results and operating data from a natural-convection water-cooled single-stage turbine

    Science.gov (United States)

    Diaguila, Anthony J; Freche, John C

    1951-01-01

    Blade-to-coolant heat-transfer data and operating data were obtained with a natural-convection water-cooled turbine over range of turbine speeds and inlet-gas temperatures. The convective coefficients were correlated by the general relation for natural-convection heat transfer. The turbine data were displaced from a theoretical equation for natural convection heat transfer in the turbulent region and from natural-convection data obtained with vertical cylinders and plates; possible disruption of natural convection circulation within the blade coolant passages was thus indicated. Comparison of non dimensional temperature-ratio parameters for the blade leading edge, midchord, and trailing edge indicated that the blade cooling effectiveness is greatest at the midchord and least at the trailing edge.

  12. Thorium in heavy water reactors

    International Nuclear Information System (INIS)

    Andersson, G.

    1984-12-01

    Advanced heavy water reactors can provide energy on a global scale beyond the foreseeable future. Their economic and safety features are promising: 1. The theoretical feasibility of the Self Sufficient Equilibrium Thorium (SSET) concept is confirmed by new calculations. Calculations show that the adjuster rod geometry used in natural uranium CANDU reactors is adequate also for SSET if the absorption in the rods is graded. 2. New fuel bundle designs can permit substantially higher power output from a CANDU reactor. The capital cost for fuel, heavy water and mechanical equipment can thereby be greatly reduced. Progress is possible with the traditional fuel material oxide, but the use of thorium metal gives much larger effects. 3. A promising long range possibility is to use pressure tanks instead of pressure tubes. Heat removal from the core is facilitated. Negative temperature and void coefficients provide inherent safety features. Refuelling under power is no longer needed if control by moderator displacement is used. Reduced quality demand on the fuel permits lower fuel costs. The neutron economy is improved by the absence of pressure and clandria tubes and also by the use of radial and axial blankets. A modular seed blanket design can reduce the Pa losses. The experience from construction of tank designs is good e.g. AAgesta, Attucha. It is now also possible to utilize technology from LWR reactors and the implementation of advanced heavy water reactors would thus be easier than HTR or LMFBR systems. (Author)

  13. Decontamination of the RA reactor heavy water system, Annex 9

    International Nuclear Information System (INIS)

    Maksimovic, Z.B.; Nikolic, R.M.; Marinkovic, M.D.; Jelic, Lj.M.

    1963-01-01

    Both stainless steel and aluminium parts of the RA reactor heavy water system system were decontaminated as well as the heavy water itself. System was contaminated with 60 Co. Decontamination factor was determined by activity measurements during distillation. Concentration of the corrosion products in the heavy water was measured by spectrochemical analysis, and found to be 0.1 - 1 mg/l. Chemical analyses of the aluminium and stainless steel surfaces showed that cobalt was adsorbed on the aluminium oxide layer. Water solution of 7%H 3 PO 4 + 2% CrO 3 was used for decontamination of the heavy water system and distillation device. This was found to be the most efficient solvent which does not affect stainless steel corrosion. Decontamination factors achieved were from 60 - 100. Decontamination results enabled determining the distribution of cobalt in the system: 10 Ci on the stainless steel parts, 50 Ci in the heavy water; and above 600 Ci on the fuel and experimental channels. Specific activity of 60 Co was calculated to be 15 Ci/g on the reactor channels, 8 Ci/g on the stainless steel parts and 3 Ci/g in the heavy water. Decontamination of the aluminium parts was not done because it was considered it could initiate corrosion. Since the efficiency of distillation is increased it was expected that permanent distillation would remove most of the activity in the reactor channels

  14. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  15. Heavy Metals Pollution on Surface Water Sources in Kaduna ...

    African Journals Online (AJOL)

    This study examine the effects of heavy metal pollutants to aquatic ecosystems and the environment by considering the role of urban, municipal, agricultural, industrial and other anthropogenic processes as sources of heavy metal pollution in surface water sources of Kaduna metropolis. Samples of the polluted water were ...

  16. Consequences of potential accidents in heavy water plants

    International Nuclear Information System (INIS)

    Croitoru, C.; Lazar, R.E.; Preda, I.A.; Dumitrescu, M.

    2002-01-01

    Heavy water plants achieve the primary isotopic concentration by H 2 O-H 2 S chemical exchange. In these plants are stored large quantities of hydrogen sulphide (high toxic, corrosive, flammable and explosive) maintained in process at relative high temperatures and pressures. It is required an assessment of risks associated with the potential accidents. The paper presents adopted model for quantitative consequences assessment in heavy water plants. Following five basic steps are used to identify the risks involved in plants operation: hazard identification, accident sequences development, H 2 S emissions calculus, dispersion analyses and consequences determination. A brief description of each step and some information from risk assessment for our heavy water pilot plant are provided. Accident magnitude, atmospheric conditions and population density in studied area were accounted for consequences calculus. (author)

  17. Heavy metals in drinking water: Occurrences, implications, and future needs in developing countries.

    Science.gov (United States)

    Chowdhury, Shakhawat; Mazumder, M A Jafar; Al-Attas, Omar; Husain, Tahir

    2016-11-01

    Heavy metals in drinking water pose a threat to human health. Populations are exposed to heavy metals primarily through water consumption, but few heavy metals can bioaccumulate in the human body (e.g., in lipids and the gastrointestinal system) and may induce cancer and other risks. To date, few thousand publications have reported various aspects of heavy metals in drinking water, including the types and quantities of metals in drinking water, their sources, factors affecting their concentrations at exposure points, human exposure, potential risks, and their removal from drinking water. Many developing countries are faced with the challenge of reducing human exposure to heavy metals, mainly due to their limited economic capacities to use advanced technologies for heavy metal removal. This paper aims to review the state of research on heavy metals in drinking water in developing countries; understand their types and variability, sources, exposure, possible health effects, and removal; and analyze the factors contributing to heavy metals in drinking water. This study identifies the current challenges in developing countries, and future research needs to reduce the levels of heavy metals in drinking water. Copyright © 2016 Elsevier B.V. All rights reserved.

  18. Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

    Science.gov (United States)

    Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

    2017-11-01

    The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

  19. Water conservation by 3 R's - case histories of Heavy Water Plants

    International Nuclear Information System (INIS)

    Agarwal, A.K.; Hiremath, S.C.

    2005-01-01

    The basics of water conservation revolve around three R's of Reduce, Recycle, and Reuse. The Heavy Water Plants are an excellent example of water savings, and these case studies will be of interest to the chemical industry. The issues involved with water conservation and re-use in different Heavy Water Plants are of different nature. In H 2 S-H 2 O process plants the water consumption has been substantially decreased as compared to the design water needs. To quote the figures HWP (Kota) was designed to consume 2280 m 3 /hr water, which included 453 m 3 /hr water as feed for deuterium extraction. Today the plant operates with only 1250 m 3 /hr water while processing 500 m 3 /hr feed; and is headed to decrease the total water consumption to 700 m 3 /hr. Similarly at HWP (Manuguru) the design had provided 5600 m 3 /hr water consumption, which is today operating with only 1750 m 3 /hr and poised to operate with 1600 m 3 /hr. The issues of water conservation in Ammonia Hydrogen exchange plants have an additional dimension since water losses mean direct loss of heavy water production. In adjoining ammonia plants deuterium shifts to steam in the reformer and shift converter, and this excess steam is condensed as rich condensate. It becomes incumbent on the fertilizer plant to maintain a tight discipline for conserving and re-using the rich condensate so that deuterium concentration in the synthesis gas is maintained. Efforts are also underway to utilize rich condensate of GSFC in the newly developed technology of water ammonia exchange at HWP (Baroda) and we are targeting 20% production gains by implementation of this scheme and with no increase in the pollution load. These case histories will be of interest to Chemical Process Industry. (author)

  20. Possibilities for reorientation the activity of heavy water plants

    International Nuclear Information System (INIS)

    Pop, F.; Croitoru, C.; Titescu, Gh.; Stefanescu, I.; Hodor, I.; Cuna, S. . E-mail of corresponding author: pop.floarea@icsi.ro; Pop, F.)

    2005-01-01

    In Romania heavy water is produced by H 2 O-H 2 S chemical exchange (GS process) and by water distillation, simultaneously working two lines. The distillation plants have high separation capacity, a distillation line being able to concentrate water from two GS lines. The paper presents data regarding possibilities to use one distillation line for oxygen 18 production, as pre-concentrates or finite products. Using a simulation program it was calculated oxygen 18 concentration in heavy water produced, maximum 18 O concentration of pre-concentrate obtained on distillation line and the separation cascade dimensions for obtain 95% 18 O, with first and second stage having same dimensions like a distillation plant from Romanian heavy water factory. Oxygen-18 separation factor is much lower than deuterium separation factor. For this reason, oxygen-18 is a very expensive product. (author)

  1. Determination of selected heavy metals in inland fresh water of ...

    African Journals Online (AJOL)

    Agadaga

    Key words: Heavy metals, freshwater, concentrations, quality, variation, distribution. ... prevalence of heavy metals in inland water of lower River. Niger drain are scarce ..... Niger waters at Ajaokuta were found to be low and within guideline.

  2. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (IIV), Part I, IZ-240-o379-1963, Vol. I, Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. Materials for irradiation are metallurgy and chemical samples. In addition to the project objectives, this volume includes technical specifications of the coolant loop head, thermal calculations, calculations of mechanical stress, antireactivity and activation of the construction materials, cost estimation, scheme of the coolant loop head, diagrams of CO 2 gas temperature, thermal neutron flux distribution, design specifications of two proposed solutions for head of low temperature coolant loop [sr

  3. Development and implementation of the heavy water program at Bruce Power

    International Nuclear Information System (INIS)

    Davloor, R.; Bourassa, C.

    2014-01-01

    Bruce Power operates 8 pressurized heavy water reactor units requiring more than 6000 mega grams (Mg) of heavy water. A Heavy Water Management Program that has been developed to administer this asset over the past 3 years. Through a corporate management system the Program provides governance, oversight and support to the stations. It is implemented through organizational structure, program and procedure documents and an information management system that provides benchmarked metrics, business intelligence and analytics for decision making and prediction. The program drives initiatives such as major maintenance activities, capital programs, detritiation strategies and ensures heavy water systems readiness for outages and rehabilitation of units. (author)

  4. Development and implementation of the heavy water program at Bruce Power

    Energy Technology Data Exchange (ETDEWEB)

    Davloor, R.; Bourassa, C., E-mail: ram.davloor@brucepower.com, E-mail: carl.bourassa@brucepower.com [Bruce Power, Tiverton, ON (Canada)

    2014-07-01

    Bruce Power operates 8 pressurized heavy water reactor units requiring more than 6000 mega grams (Mg) of heavy water. A Heavy Water Management Program that has been developed to administer this asset over the past 3 years. Through a corporate management system the Program provides governance, oversight and support to the stations. It is implemented through organizational structure, program and procedure documents and an information management system that provides benchmarked metrics, business intelligence and analytics for decision making and prediction. The program drives initiatives such as major maintenance activities, capital programs, detritiation strategies and ensures heavy water systems readiness for outages and rehabilitation of units. (author)

  5. Leak detection device for reactor coolant

    International Nuclear Information System (INIS)

    Oshima, Koichiro.

    1990-01-01

    In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

  6. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

  7. Measurement of vibrations in the primary coolant circuit and in the vertical experimental channel of the RA reactor

    International Nuclear Information System (INIS)

    Ristic, B.; Rakic, R.; Milosevic, M.; Jerkovic, M.

    1966-01-01

    Full text: Beginning of the work dates from 1962 with the initial objective: study of the wear-out of the bearings of the centrifugal pumps in the heavy water system. It has been expected that the increase of wear-out would initiate increase of vibration amplitudes and noise. During further study the initial task was broadened to other fields, mainly appearance of material fatigue in components of the heavy water coolant system. During operation mechanical energy is generated due to non existing equilibrium of the pump rotor, wear-out of the bearing, turbulence in the pump, cavitation process and pulsation of the operating environment. This energy is transformed into noise and vibration energy which is spread through surrounding walls and pipes causing noise finally. Obtained results were only qualitatively tested at present. For quantitative testing it would be necessary to obtain data about the material, in addition to the diagrams obtained by measurements. It would be possible to calculate the fatigue of the material at measuring points as well as estimation of the time when material fatigue would become critical [sr

  8. Undermoderated spectrum MOX core study. Pressurized water-type breeder

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi; Komano, Yasuo

    1998-01-01

    The purpose of this development is the advance of the PWR core. The conversion ratio (the breeding ratio) are examined. In the case of heavy water as the coolant, the breeding ratio can be achieved about 1.1 with using a hexagonal lattice and space about 1 mm assembly fuel. In the case of light water coolant, the breeding ratio becomes about 1.0, using a hexagonal lattice and fuel space about 0.5 mm fuel assembly. Here, it reports on the situation of the examination such as the nucleus design of core, the design of fuel assembly, the heat hydraulics design of the core, the structure design and so on. (author)

  9. Possibilities for reorientation of activity in Heavy Water Plants

    International Nuclear Information System (INIS)

    Pop, F.; Croitoru, C.; Titescu, Gh.; Stefanescu, I.; Hodor, I.; Cuna, S.

    2004-01-01

    In Romania heavy water is produced by H 2 O-H 2 S chemical exchange (GS process) and by water distillation, in two lines working simultaneously. The distillation plants have high separation capacity, a distillation line being able to concentrate water from two GS lines. The paper presents data regarding possibilities to use one distillation line for oxygen - 18 production, as pre-concentrates or finite products. A simulation program was used to calculate the oxygen - 18 concentration in the heavy water produced, maximum 18 O concentration of pre-concentrate obtained on distillation line and the separation cascade sizes to obtain 95% 18 O, with first and second stage having the same sizes like the distillation plant from the Romanian heavy water factory. Oxygen-18 separation factor is much lower than deuterium separation factor. For this reason, oxygen-18 is a very expensive product. (authors)

  10. Study of the heavy water regeneration processes; Studija procesa za regeneraciju teske vode

    Energy Technology Data Exchange (ETDEWEB)

    Cavcic, E [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1965-11-15

    Experience derived from heavy water reactor operation showed degradation and dilution of heavy water to be inevitable depends on the type of reactor. Dilution of heavy water during operation of the RA and the RB reactors is shown in this report. Principles and procedures of heavy water regeneration by electrolysis, fractional distillation, cleaning, prevention of tritium contamination are described as well as separation columns.

  11. Effects of Specific Fuel Consumption and Exhaust Emissions of Four Stroke Diesel Engine with CuO/Water Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Senthilraja S.

    2017-03-01

    Full Text Available This article reports the effects of CuO/water based coolant on specific fuel consumption and exhaust emissions of four stroke single cylinder diesel engine. The CuO nanoparticles of 27 nm were used to prepare the nanofluid-based engine coolant. Three different volume concentrations (i.e 0.05%, 0.1%, and 0.2% of CuO/water nanofluids were prepared by using two-step method. The purpose of this study is to investigate the exhaust emissions (NOx, exhaust gas temperature and specific fuel consumption under different load conditions with CuO/water nanofluid. After a series of experiments, it was observed that the CuO/water nanofluids, even at low volume concentrations, have a significant influence on exhaust emissions. The experimental results revealed that, at full load condition, the specific fuel consumption was reduced by 8.6%, 15.1% and 21.1% for the addition of 0.05%, 0.1% and 0.2% CuO nanoparticles with water, respectively. Also, the emission tests were concluded that 881 ppm, 853 ppm and 833 ppm of NOx emissions were observed at high load with 0.05%, 0.1% and 0.2% volume concentrations of CuO/water nanofluids, respectively.

  12. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

    Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

  13. Energy conservation and management strategies in Heavy Water Plants

    International Nuclear Information System (INIS)

    Kamath, H.S.

    2002-01-01

    In the competitive industrial environment it is essential that cost of the product is kept at the minimum possible. Energy conservation is an important aspect in achieving this as energy is one of the key recourses for growth and survival of industry. The process of heavy water production being very complex and energy intensive, Heavy Water board has given a focussed attention for initiating various measures for reducing the specific energy consumption in all the plants. The initiative resulted in substantial reduction in specific energy consumption and brought in savings in cost. The cumulative reduction of specific energy consumption has been over 30% over the last seven years and the total savings for the last three years on account of the same has been about Rs. 190 crore. The paper describes the strategies adopted in the heavy water plants for effecting the above achievements. The paper covers the details of some of the energy saving schemes carried out at different heavy water plants through case studies. The case studies of schemes implemented at HWPs are general in nature and is applicable for any other industry. The case studies cover the modifications with re-optimisation of the process parameters, improvements effected in utility units like refrigeration and cooling water systems, improvements in captive power plant cycle and improved recycle scheme for water leading to reduced consumptions. The paper also mentions the innovative ammonia absorption refrigeration with improved coefficient of performance and HWB's efforts in development of the system as an integrated unit of the ammonia water deuterium exchange process for heavy water production. HWB also has taken up R and D on various other schemes for improvements in energy consumption for future activities covering utilisation of low grade energy for generation of refrigeration. (author)

  14. Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

    International Nuclear Information System (INIS)

    Johnson, K.R.

    1981-01-01

    A rigorous semi-empirical approach was undertaken to model the response of an ultrasonic level device (ULD) for application to in-core coolant detection in Pressurized Water Reactors (PWRs). An equation is derived for the torsional wave velocity v/sub t phi/ in the ULD. Existing data reduction techniques were analyzed and compared to results from use of the derived equation. Both methods yield liquid level measurements with errors of approx. 5%. A sensitivity study on probe performance at reactor conditions predicts reduced level responsivity from data at lower temperatures

  15. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    zinc compounds (mainly borates) were observed at the heatable zircaloy surfaces and characterized in detail during the heating-up to several coolant temperatures. As a strict consequence of their proven influence on heat removal and coolant flow behavior in the PWR core, preventive water-chemical methods were defined and tested.

  16. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  17. Return momentum effect on reactor coolant water level distribution during mid-loop conditions

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Yang, Jae Young; Park, Goon Cherl

    2001-01-01

    An accurate prediction of the Reactor Coolant System( RCS) water level is of importance in the determination of the allowable operating range to ensure safety during mid-loop operations. However, complex hydrualic phenomena induced by the Shutdown Cooling System (SCS) return momentum causes different water levels from those in the loop where the water level indicators are located. This was apparently observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level distribution, a model using a one-dimensional momentum and energy conservation for cylindrical channel, hydraulic jump in operating cold leg, water level build-up at the Reactor Vessel (RV) inlet nozzle, Bernoulli constant in downcomer region, and total water volume conservation has been developed. The model predicts the RCS water levels at various RCS locations during the mid-loop conditions and the calculation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs, in conjuction with the pressure drop throughout the RCS, is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

  18. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  19. Conceptual design of a large heavy water reactor for US siting

    International Nuclear Information System (INIS)

    Shapiro, N.L.; Jesick, J.F.

    1979-09-01

    Information is presented concerning fuel management and safety and licensing assessment of the pressurized heavy water reactor; and commercial introduction of the pressurized heavy water reactor in the United States

  20. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  1. 76 FR 52994 - Application for a License To Export Heavy Water

    Science.gov (United States)

    2011-08-24

    ... NUCLEAR REGULATORY COMMISSION Application for a License To Export Heavy Water Pursuant to 10 CFR... (liters). producing an active water). pharmaceutical ingredient known as CTP-499, which incorporates heavy water as the source of deuterium to achieve the hydrogen-deuterium exchange. November 30, 2010 December...

  2. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  3. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  4. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  5. Parametric study of postulated reactivity transients due to ingress of heavy water from the reflector tank into the converted core of APSARA reactor

    International Nuclear Information System (INIS)

    Sankaranarayanan, S.

    2004-01-01

    Research reactors in the power range 5-10 MW with useable neutron flux values >1.OE+14 n/sqcm/sec can be constructed using LEU fuel with light water for neutron moderation and fuel cooling. In order to obtain a large irradiation volume, a heavy water reflector is used where fairly high neutron flux levels can be obtained. A prototype LEU fuelled 5/10 MW reactor design has been developed in the Bhabha Atomic Research Centre in Trombay. Work is on hand to carry out technology simulation of this reactor design by converting the pool type reactor APSARA in BARC. Presently the Apsara reactor uses MTh type high enriched U-Al alloy plate type fuel loaded in a 7x7 grid with a square lattice pitch of 76.8 mm. The reactor has three control-scram-shut off rods and one regulating control rod. In the first phase of the simulation studies, it is proposed to use the existing high enriched uranium fuel in a modified core with 37 positions arranged with a square lattice pitch of 84.8 mm, surrounded by a 50 cm thick heavy water reflector. Subsequently the converted core will use plate-type low enriched uranium suicide fuel. One of the accident scenarios postulated for the safety evaluation of the modified APSARA reactor is the reactivity transient due to the ingress of heavy water into the core through a small sized rupture in the aluminium wall of the reflector tank. Parametric analyses were done for the safety evaluation of modified Apsara reactor, for postulated leak of heavy water into the core from the reflector tank. A simplified computer code REDYN, based on point model reactor kinetics with one effective group of delayed neutrons is used for the analyses. Results of several parametric cases used in the study show that it is possible to contain the consequences of this type of reactivity transient within acceptable fuel and coolant thermal safety limits

  6. Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

    International Nuclear Information System (INIS)

    Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

    2010-01-01

    Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

  7. Performance Analysis of Thermoelectric Based Automotive Waste Heat Recovery System with Nanofluid Coolant

    Directory of Open Access Journals (Sweden)

    Zhi Li

    2017-09-01

    Full Text Available Output performance of a thermoelectric-based automotive waste heat recovery system with a nanofluid coolant is analyzed in this study. Comparison between Cu-Ethylene glycol (Cu-EG nanofluid coolant and ethylene glycol with water (EG-W coolant under equal mass flow rate indicates that Cu-EG nanofluid as a coolant can effectively improve power output and thermoelectric conversion efficiency for the system. Power output enhancement for a 3% concentration of nanofluid is 2.5–8 W (12.65–13.95% compared to EG-Water when inlet temperature of exhaust varies within 500–710 K. The increase of nanofluid concentration within a realizable range (6% has positive effect on output performance of the system. Study on the relationship between total area of thermoelectric modules (TEMs and output performance of the system indicates that optimal total area of TEMs exists for maximizing output performance of the system. Cu-EG nanofluid as coolant can decrease optimal total area of TEMs compared with EG-W, which will bring significant advantages for the optimization and arrangement of TEMs whether the system space is sufficient or not. Moreover, power output enhancement under Cu-EG nanofluid coolant is larger than that of EG-W coolant due to the increase of hot side heat transfer coefficient of TEMs.

  8. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  9. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  10. Spatiotemporal Analysis of Heavy Metal Water Pollution in Transitional China

    Directory of Open Access Journals (Sweden)

    Huixuan Li

    2015-07-01

    Full Text Available China’s socioeconomic transitions have dramatically accelerated its economic growth in last three decades, but also companioned with continuous environmental degradation. This study will advance the knowledge of heavy metal water pollution in China from a spatial–temporal perspective. Specifically, this study addressed the following: (1 spatial patterns of heavy metal water pollution levels were analyzed using data of prefecture-level cities from 2004 to 2011; and (2 spatial statistical methods were used to examine the underlying socioeconomic and physical factors behind water pollution including socioeconomic transitions (industrialization, urbanization, globalization and economic development, and environmental characteristic (natural resources, hydrology and vegetation coverage. The results show that only Cr pollution levels increased over the years. The individual pollution levels of the other four heavy metals, As, Cd, Hg, and Pb, declined. High heavy metal water pollution levels are closely associated with both anthropogenic activities and physical environments, in particular abundant mineral resources and industrialization prosperity. On the other hand, economic development and urbanization play important roles in controlling water pollution problems. The analytical findings will provide valuable information for policy-makers to initiate and adjust protocols and strategies for protecting water sources and controlling water pollution; thus improving the quality of living environments.

  11. Extending the product variety at ROMAG-PROD Heavy Water Plant

    International Nuclear Information System (INIS)

    Preda, M.; Patrascu, M.; Achimescu, D.; Stroia, A.

    2004-01-01

    Full text: Having in mind that the prospects of operating the ROMAG-PROD Heavy Water Plant are conditioned by both the heavy water market demand and the wear of the equipment which is exposed to hydrogen sulfide-induced corrosion, some possibilities were considered to extend the assortment of products, the production of which could ensure the plant's operation on long term. The proposals here refer to promoting the efficient production of oxygen-isotope-based products which would optimize maximally the exploit of available raw materials, supply and utilities of the ROMAG compound. The market manifests a significant demand of water enriched in the 18 O isotope up to 95-97% purity that is used in Positron Emission Tomography (PET). This oxygen isotope is also used as a labelling agent in studies of reaction mechanisms and paleo-climatologic studies as well. Some research evidenced the superconducting properties of some oxygen compounds containing the 18 O isotope. The isotope 17 O has applications in Nuclear Magnetic Resonance (NMR) as being the sole oxygen isotope endowed with a nuclear magnetic moment. On the other hand, it was found that although the 16 O isotope has a natural abundance of 99.8%, applications exist that require the absolute purity of this isotope i.e. the elimination of the other oxygen isotopes as is the case of fission reactors with Plutonium dioxide as nuclear fuel. The methods applied on industrial scale for enriching the oxygen isotopes are based on distillation of some oxygen compounds such as water and nitrogen monoxide. The possibility of a supplementary distillation of the heavy water at a distillation line of ROMAG-PROD Heavy Water Plant was considered in order to enrich the heavy water in the 17 O and 18 O isotopes up to an upper limit of 2-5% for 18 O. Obtaining the heavy isotopes of oxygen by distillation of heavy water is characterized by several aspects as the following ones: a high specific consumption of steam due to both the low

  12. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (I-IV), Part II, IZ-240-0379-1963, Vol. II Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. maximum generated heat in the heads of the coolant loop is 10500 kcal/h and minimum generated heat is 1500 kcal/h. The loops are cooled by CO 2 gas, coolant flow is 420 kg/h, and the pressure is 4.5 atu. There is a need to design and construct the secondary coolant loop for the low temperature coolant loop. This volume includes technical specifications of the secondary CO 2 loop with instructions for construction and testing; needed calculations; specification of materials; cost estimation for materials, equipment and construction; and graphical documentation [sr

  13. 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

  14. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  15. Fuel-coolant interaction visualization test for in-vessel corium retention external reactor vessel cooling (IVR-ERVC) condition

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Hong, Seong Ho; Song, Jin Ho; Hong, Seong Wan [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-12-15

    A visualization test of the fuel-coolant interaction in the Test for Real cOrium Interaction with water (TROI) test facility was carried out. To experimentally simulate the In-Vessel corium Retention (IVR)- External Reactor Vessel Cooling (ERVC) conditions, prototypic corium was released directly into the coolant water without a free fall in a gas phase before making contact with the coolant. Corium (34.39 kg) consisting of uranium oxide and zirconium oxide with a weight ratio of 8:2 was superheated, and 22.54 kg of the 34.39 kg corium was passed through water contained in a transparent interaction vessel. An image of the corium jet behavior in the coolant was taken by a high-speed camera every millisecond. Thermocouple junctions installed in the vertical direction of the coolant were cut sequentially by the falling corium jet. It was clearly observed that the visualization image of the corium jet taken during the fuel-coolant interaction corresponded with the temperature variations in the direction of the falling melt. The corium penetrated through the coolant, and the jet leading edge velocity was 2.0 m/s. Debris smaller than 1 mm was 15% of the total weight of the debris collected after a fuel-coolant interaction test, and the mass median diameter was 2.9 mm.

  16. Heavy water technology and its contribution to energy sustainability

    International Nuclear Information System (INIS)

    MacDiarmid, H.; Alizadeh, A.; Hopwood, J.; Duffey, R.

    2009-01-01

    Full text: As the global nuclear industry expands several markets are exploring avenues and technologies to underpin energy security. Heavy water reactors are the most versatile power reactors in the world. They have the potential to extend resource utilization significantly, to allow countries with developing industrial infrastructures access to clean and abundant energy, and to destroy long-lived nuclear waste. These benefits are available by choosing from an array of possible fuel cycles. Several factors, including Canada's early focus on heavy-water technology, limited heavy-industry infrastructure at the time, and a desire for both technological autonomy and energy self-sufficiency, contributed to the creation of the first commercial heavy water reactor in 1962. With the maturation of the industry, the unique design features of the now-familiar product-on-power refuelling, high neutron economy, and simple fuel design-make possible the realization of its potential fuel-cycle versatility. As resource constrains apply pressure on world markets, the feasibility of these options have become more attractive and closer to entering widespread commercial application

  17. HEAVY METALS AS UNWANTED COMPONENTS OF BACKWASH WATER DERIVED FROM GROUNDWATER TREATMENT

    Directory of Open Access Journals (Sweden)

    Robert Nowak

    2016-06-01

    Full Text Available The paper presents some aspects of the problem of heavy metals presence in wastewater and sewage sludge from water treatment. In the first part, issues on quality of wastewaters and sludge produced during water treatment along with actions aimed at the neutralization of such wastes, were discussed. Subsequent parts of the work present the example of 12 groundwater treatment stations in a particular municipality, and the problem of backwash water quality, in particular, heavy metals contents. The analysis covered a period of three years: 2013, 2014, and 2015. The authors, using the discussed examples, have shown that besides hydrated iron and manganese oxides, also other toxic contaminants can be present in backwash water from groundwater treatment. In particular, the qualitative analysis of the backwash water revealed the presence of heavy metals, mainly zinc. The test results for backwash water were compared with those of filtrate qualitative assessment, wherein the heavy metals were not found. This fact indicated the metal retention in the filter bed and their unsustainable immobilization resulting in penetration of heavy metals from deposit to the backwash water along with other impurities, mainly iron and manganese oxides. The main conclusion from the study is to demonstrate the need for constant monitoring of the backwash water quality, including the presence of toxic heavy metals. This is also important because of the requirement to minimize the negative environmental impact of wastes generated during the water treatment process.

  18. "Periodic-table-style" paper device for monitoring heavy metals in water.

    Science.gov (United States)

    Li, Miaosi; Cao, Rong; Nilghaz, Azadeh; Guan, Liyun; Zhang, Xiwang; Shen, Wei

    2015-03-03

    If a paper-based analytical device (μ-PAD) could be made by printing indicators for detection of heavy metals in chemical symbols of the metals in a style of the periodic table of elements, it could be possible for such μ-PAD to report the presence and the safety level of heavy metal ions in water simultaneously and by text message. This device would be able to provide easy solutions to field-based monitoring of heavy metals in industrial wastewater discharges and in irrigating and drinking water. Text-reporting could promptly inform even nonprofessional users of the water quality. This work presents a proof of concept study of this idea. Cu(II), Ni(II), and Cr(VI) were chosen to demonstrate the feasibility, specificity, and reliability of paper-based text-reporting devices for monitoring heavy metals in water.

  19. Canned motor pumps at Heavy Water Project, Baroda

    International Nuclear Information System (INIS)

    Batra, R.K.; Waishampayan, S.C.

    1981-01-01

    Pumps to be used in heavy water plants must be reliable and should require negligible maintenance, because most of them are totally unapproachable under normal circumstances. Canned motor pumps fulfil these requirements. Their design features are described briefly. The details of: (1) the pumps in the isotopic exchange tower and (2) pumps for liquid ammonia and catalyst are given. Problems faced during commissioning of such pumps in Baroda Heavy Water Project were bulging of rotors of tower pumps, bulging of stators, jamming and failure of bearings. Solution of these problems is described. (M.G.B.)

  20. Contact condensation effects in the main coolant pipe

    International Nuclear Information System (INIS)

    Haefner, W.; Fischer, K.

    1990-01-01

    Contact condensation effects may occur in a pressurized water reactor (PWR) after a loss of coolant accident (LOCA) when emergency core cooling (ECC) water is injected contact with escaping steam which is generated within the core. The condensation which takes place may cause a sudden depressurization leading to the formation of water slugs. The interaction between the transient condensation and the inertia of the flow may also result in large amplitude flow and pressure oscillations. These contact condensation effects are of great importance for the mass flow distribution and the coolant water supply to the reactor core. To examine those complex processes, large computer codes are necessary. The development and verification of analytical models requires greatly simplified flow boundary conditions from experiments and a sufficiently large base of experimental data. Separate models have been developed for interfacial exchange of mass, momentum and energy with respect to the associated flow regime. Therefore, an adequate description of the condensation process requires the modeling of two different topics: the prediction of the flow regime and the calculation of the interfacial exchange. (author)

  1. Corrosion fatigue studies on F82H mod. martensitic steel in reducing water coolant environments

    Energy Technology Data Exchange (ETDEWEB)

    Maday, M F; Masci, A [ENEA, Casaccia (Italy). Centro Ricerche Energia

    1998-03-01

    Load-controlled low cycle fatigue tests have been carried out on F82H martensitic steel in 240degC oxygen-free water with and without dissolved hydrogen, in order to simulate realistic coolant boundary conditions to be approached in DEMO. It was found that water independently of its hydrogen content, determined the same fatigue life reduction compared to the base-line air results. Water cracks exhibited in their first propagation stages similar fracture morphologies which were completely missing on the air cracks, and were attributed to the action of an environment related component. Lowering frequency gave rise to an increase in F82H fatigue lifetimes without any change in cracking mode in air, and to fatigue life reduction by microvoid coalescence alone in water. The data were discussed in terms of (i) frequency dependent concurrent processes for crack initiation and (ii) frequency-dependent competitive mechanisms for crack propagation induced by cathodic hydrogen from F82H corrosion. (author)

  2. Thermal neutron standard fields with the KUR heavy water facility

    International Nuclear Information System (INIS)

    Kanda, K.; Kobayashi, K.; Shibata, T.

    1978-01-01

    A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type, 5 MW) yields pure thermal neutrons in the Maxwellian distribution. The facility is faced to the core of KUR and it contains about 2 tons of heavy water. The thickness of the layer is about 140 cm. The neutron spectrum was measured with the time of flight technique using a fast chopper. The measured spectrum was in good agreement with the Maxwellian distribution in all energy region for thermal neutrons. The neutron temperature was slightly higher than the heavy water temperature. The contamination of epithermal and fast neutrons caused by photo-neutrons of the γ-n reaction of heavy water was very small. The maximum intensity of thermal neutrons is 3x10 11 n/cm 2 sec. When the bismuth scatterer is attached, the gamma rays contamination is eliminated by the ratio of 0.05 of gamma rays to neutrons in rem. This standard neutron field has been used for such experiments as thermal neutron cross section measurement, detector calibration, activation analysis, biomedical purposes etc. (author)

  3. 1000 tones of heavy water produced at ROMAG PROD, Drobeta-Turnu Severin

    International Nuclear Information System (INIS)

    2001-01-01

    On May 25, 2001 the heavy water plant ROMAG PROD at Drobeta-Turnu Severin recorded the production of the 1000-th tone of nuclear purity heavy water. The heavy water plant ROMAG PROD makes use of a technology based on the results of isotopic deuterium separation research carried out at the Research and Design Institutes of Cluj, Craiova, Pitesti and Ploiesti during 1957-1970 and the separation technology tested at Ramnicu-Valcea pilot plant (at present the Cryogenics and Isotope Separation Institute). The first investments at ROMAG PROD were made in 1979 and on July 17, 1988 was produced the first amount of heavy water at the required parameters for CANDU type nuclear reactors. The period between 1990-1992 was dedicated to the project completion, upgrading the technological facilities and retrofitting the environmental protection and monitoring systems. Production was resumed in 1992. The first 500 t of heavy water required for the Cernavoda NPP first reactor operation were produced by summer 1997. The additional amount of 500 t of heavy water was produced between 1997-2001. ROMAG PROD obtained the ISO 9001/2001 certificate for the quality management system, the ISO 14001/1997 certificate for the environmental management system and the new environmental permit

  4. 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

  5. Heat exchangers in heavy water reactor systems

    International Nuclear Information System (INIS)

    Mehta, S.K.

    1988-01-01

    Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

  6. Environmental assessment of ground water pollution by heavy ...

    African Journals Online (AJOL)

    The aim of this study was to investigate the relationship between the concentrations of heavy metals in well water and bioaccumulation of the most abundant metals in chicken tissues in some areas in the province of Mecca Almokaramah, Saudi Arabia. Among the heavy metals (Cd, Zn, Cr, Mn, Cu Hg, Pb and Ni) studied, ...

  7. Evaluation of primary coolant pH operation methods for the domestic PWRs

    International Nuclear Information System (INIS)

    Paek, Seung Woo; Na, Jung Won; Kim, Yong Eak; Bae, Jae Heum

    1992-01-01

    Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of-core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed. (Author)

  8. Eddy current monitoring of spacers in coolant channel assemblies of nuclear reactor

    International Nuclear Information System (INIS)

    Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.; Vijayaraghavan, R.

    1993-01-01

    An eddy current testing method has been standardised for monitoring spacer springs which are used in coolant channel assemblies of pressurised heavy water nuclear reactors (PHWRs). The standard bobbin coil probe used for monitoring the spacer spring detects only the location but does not monitor the tilt orientation and tilt angle of a tilted spacer spring. The knowledge of location along with the tilt orientation of the spacer spring greatly improves the performance of repositioning methods. A modified probe with angular windings has been developed in laboratory tests for monitoring the location as well as the tilt orientation of the spacer springs. Experimental results are presented showing excellent performance of the modified probe in monitoring the exact location as well as tilt orientation of a spacer spring. The modified probe has also been used successfully in the field during repositioning of spacer springs in PHWRs before commissioning. (Author)

  9. Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

    2012-07-01

    In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

  10. Operating experiences on ammonia water exchange system at Heavy Water Plant, Talcher (paper No. 6.12)

    International Nuclear Information System (INIS)

    Venkat Ram, D.; Sharma, A.K.

    1992-01-01

    The Heavy Water Plant at Talcher employs bithermal ammonia hydrogen exchange process for the production of heavy water. The paper describes about the existing ammonia water exchange column, its start-up, operating experience and the problems encountered in operation of the column. The operating experiences gained and the data collected over the last few years can be utilised for design and operation of new ammonia water exchange column. (V.R). 2 figs

  11. Heavy metal partitioning of suspended particulate matter-water and sediment-water in the Yangtze Estuary.

    Science.gov (United States)

    Feng, Chenghong; Guo, Xiaoyu; Yin, Su; Tian, Chenhao; Li, Yangyang; Shen, Zhenyao

    2017-10-01

    The partitioning of ten heavy metals (As, Cd, Co, Cr, Cu, Hg, Ni, Pb, Sb, and Zn) between the water, suspended particulate matter (SPM), and sediments in seven channel sections during three hydrologic seasons in the Yangtze Estuary was comprehensively investigated. Special attention was paid to the role of tides, influential factors (concentrations of SPM and dissolved organic carbon, and particle size), and heavy metal speciation. The SPM-water and sediment-water partition coefficients (K p ) of the heavy metals exhibited similar changes along the channel sections, though the former were larger throughout the estuary. Because of the higher salinity, the K p values of most of the metals were higher in the north branch than in the south branch. The K p values of Cd, Co, and As generally decreased from the wet season to the dry season. Both the diagonal line method and paired samples t-test showed that no specific phase transfer of heavy metals existed during the flood and ebb tides, but the sediment-water K p was more concentrated for the diagonal line method, owing to the relatively smaller tidal influences on the sediment. The partition coefficients (especially the K p for SPM-water) had negative correlations with the dissolved organic carbon (DOC) but positive correlations were noted with the particle size for most of the heavy metals in sediment. Two types of significant correlations were observed between K p and metal speciation (i.e., exchangeable, carbonate, reducible, organic, and residual fractions), which can be used to identify the dominant phase-partition mechanisms (e.g., adsorption or desorption) of heavy metals. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. FILM-30: A Heat Transfer Properties Code for Water Coolant

    International Nuclear Information System (INIS)

    MARSHALL, THERON D.

    2001-01-01

    A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating

  13. Measurement data of cesium 137 yields in primary coolant of an in-pile water loop in fission products release experiment

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo

    1979-03-01

    Series of fuel rods (UO 2 pellets sheathed with stainless steel) having an artificial pinhole were irradiated in the in-pile test section of water loop JMTR OWL-1. Presented are the results of measurements of cesium 137 yields in primary coolant of OWL-1 from 1975 to 1978. (author)

  14. Analytical modelling and study of the stability characteristics of the Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Nayak, A.K.; Vijayan, P.K.; Saha, D.

    2000-04-01

    An analytical model has been developed to study the thermohydraulic and neutronic-coupled density-wave instability in the Indian Advanced Heavy Water Reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the design configuration considered may experience both Ledinegg and Type I and Type II density-wave instabilities depending on the operating condition. Some methods of suppressing these instabilities were found out. In addition, it was found that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have strong influence on the Type I and Type II instabilities. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (author)

  15. Tritium in water monitor for measurement of tritium activity in the process water

    International Nuclear Information System (INIS)

    Rathnakaran, M.; Ravetkar, R.M.; Abani, M.C.; Mehta, S.K.

    1999-01-01

    This paper presents the evaluation of a tritium in water monitor for measurement of tritium activity in the secondary coolant in pressurised heavy water reactor used for power generation. For this purpose it uses a plastic scintillator flow cell detector in a continuous on-line mode. It is observed that the sensitivity of the system depends on the transparency of the detector, which gradually reduces with use because of the collection of dirt around the scintillator. A simple type of sample conditioner based on polypropylene candle filter and filter paper is developed and installed at RAPS along with tritium in water monitor. The functioning of this system is reported here. (author)

  16. Fact and fiction in ECP measurement and control in boiling water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2005-01-01

    A review is presented of various electrochemical potentials, including the electrochemical corrosion potential (ECP), that are used in the mitigation of stress corrosion cracking in the primary coolant circuits of boiling water reactors (BWRs). Attention is paid to carefully defining each potential in terms of fundamental electrochemical concepts, so as to counter the confusion that has arisen due to the misuse of previously accepted terminology. A brief discussion is also included of reference electrodes and it is shown on the basis of experimental data that the use of a platinum redox sensor as a reference electrode in the monitoring of ECP in BWR primary coolant circuits is inappropriate and should be discouraged. If platinum is used as a reference electrode, because of extenuating circumstances (e.g., potential measurements in high dose regions in a reactor core), the onus must be placed on the user to demonstrate quantitatively that the electrode behaves as an equilibrium electrode under the specified conditions and/or that its potential is invariant with changes in the independent variables of the system. Preferably, a means should also be demonstrated of transferring the measured potential to the standard hydrogen electrode (SHE) scale. (orig.)

  17. Change of deuterium volume content in heavy water during carbon dioxide dissolution in it

    International Nuclear Information System (INIS)

    Efimova, T.I.; Kapitanov, V.F.; Levchenko, G.V.

    1985-01-01

    Carbon dioxide solution density in heavy water at increased temperature and pressure is measured and the influence of carbon dioxide solubility in heavy water on volumetric content of deuterium in it is determined. Investigations were conducted in the temperature range of 303-473 K and pressure range of 3-20 MPa by the autoclave method. Volumetric content of deuterium in heavy water decreases sufficiently with CO 2 dissolved in it in comparison with pure D 2 O under the similar conditions, and this decrease becomes more sufficient with the pressure increase. With the temperature increase the volumetric content of deuterium both for heavy water and for saturated carbon solution in heavy water decreases

  18. Continuous control of pH value and chloride concentration in a water coolant of nuclear reactors

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Krasnoperov, V.M.; Fokina, K.G.; Vilkov, N.Ya.

    1975-01-01

    Potentiometry method with the use of flowing cells with two identical electrodes is the simplest and most safe for continuous pH value and chloride control in nuclear reactor circulating circuits. The constant potential on the comparison electrode may be provided by supplying the analyzed solution to it through the ion resin filter of mixed operation. The pos--sibility of a continuous pH value monitoring in a flowing cell with two glass electrodes in parallel is considered. To monitor clorides a cell with two porous chlorine-silver electrodes positioned in series is used. The cells of the design described are shown to be workable in water simulating coolants for water-cooled reactors

  19. Removal of gadolinium nitrate from heavy water

    Energy Technology Data Exchange (ETDEWEB)

    Wilde, E.W.

    2000-03-22

    Work was conducted to develop a cost-effective process to purify 181 55-gallon drums containing spent heavy water moderator (D2O) contaminated with high concentrations of gadolinium nitrate, a chemical used as a neutron poison during former nuclear reactor operations at the Savannah River Site (SRS). These drums also contain low level radioactive contamination, including tritium, which complicates treatment options. Presently, the drums of degraded moderator are being stored on site. It was suggested that a process utilizing biological mechanisms could potentially lower the total cost of heavy water purification by allowing the use of smaller equipment with less product loss and a reduction in the quantity of secondary waste materials produced by the current baseline process (ion exchange).

  20. Selected bibliography on deuterium isotope effects and heavy water

    International Nuclear Information System (INIS)

    Dave, S.M.; Donde, M.M.

    1983-01-01

    In recent years, there has been a great deal of interest in using deuterium and heavy water not only in nuclear industry but also in various fields of basic as well as applied research in physics, chemistry and biology. As a result, the literature is being enriched with a large number of research papers and technical reports published each year. Thus, to enable the scientists to have an easy reference to these works, an endeavour has been made in this selected bibliography, to enlist the publications related to these fields. Since the interest is concerned mainly with heavy water production processes, deuterium isotope effects etc., several aspects (e.g. nuclear) of deuterium have not been covered here. The material in this bibliography which cites 2388 references has been classified under six broad headings, viz. (1) Production of heavy water, (2) Study of deuterium isotope effects, (3) Analysis and Properties of heavy water, (4) Laser Separation of deuterium, (5) Isotopic exchange reactions, and (6) Miscellaneous. The sources of information used for this compilation are chemical abstracts, nuclear science abstracts, INIS Atomindex and also some scattered search through journals and reports available in the B.A.R.C. library. However, in spite of sincere attempts for a wide coverage, no claim is being made towards the exhaustiveness of this bibliography. (author)

  1. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  2. Eliminating Heavy Metals from Water with NanoSheet Minerals as Adsorbents

    Directory of Open Access Journals (Sweden)

    Shaoxian Song

    2017-12-01

    Full Text Available Heavy metals usually referred to those with atomic weights ranging from 63.5 to 200.6. Because of natural-mineral dissolution and human activities such as mining, pesticides, fertilizer, metal planting and batteries manufacture, etc., these heavy metals, including zinc, copper, mercury, lead, cadmium and chromium have been excessively released into water courses, like underground water, lake and river, etc. The ingestion of the heavy metals-contaminated water would raise serious health problems to human beings even at a low concentration. For instance, lead can bring human beings about barrier to the normal function of kidney, liver and reproductive system, while zinc can cause stomach cramps, skin irritations, vomiting and anemia. Mercury is a horrible neurotoxin that may result in damages to the central nervous system, dysfunction of pulmonary and kidney, chest and dyspnea. Chromium (VI has been proved can cause many diseases ranging from general skin irritation to severe lung carcinoma. Accordingly, the World Health Organization announced the maximum contaminant levels (MCL for the heavy metals in drinking water. There are numerous processes for eliminating heavy metals from water in order to provide citizens safe drinking water, including precipitation, adsorption, ion exchange, membrane separation and biological treatment, etc. Adsorption is considered as a potential process for deeply removing heavy metals, in which the selection of adsorbents plays a predominant role. Nano-sheet minerals as the adsorbents are currently the hottest researches in the field. They are obtained from layered minerals, such as montmorillonite, graphite and molybdenite, through the processing of intercalation, electrochemical and mechanical exfoliation, etc. Nano-sheet minerals are featured by their large specific surface area, relatively low costs and active adsorbing sites, leading to be effective and potential adsorbents for heavy metals removal from water

  3. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. 1.2. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1981), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1986), which are superseded by this new Safety Guide. 1.3. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1981 and 1986, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2000, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included

  4. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

    2004-04-01

    The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

  5. Contribution to the optimization of the chemical and radiochemical purification of pressurized water nuclear power plants primary coolant

    International Nuclear Information System (INIS)

    Elain, L.

    2004-12-01

    The primary coolant of pressurised water reactors is permanently purified thanks to a device, composed of filters and the demineralizers furnished with ion exchange resins (IER), located in the chemical and volume control system (CVCS). The study of the retention mechanisms of the radio-contaminants by the IER implies, initially, to know the speciation of the primary coolant percolant through the demineralizers. Calculations of theoretical speciation of the primary coolant were carried out on the basis of known composition of the primary coolant and thanks to the use of an adapted chemical speciation code. A complementary study, dedicated to silver behaviour, considered badly extracted, suggests metallic aggregates existence generated by the radiolytic reduction of the Ag + ions. An analysis of the purification curves of the elements Ni, Fe, Co, Cr, Mn, Sb and their principal radionuclides, relating to the cold shutdown of Fessenheim 1-cycle 20 and Tricastin 2-cycle 21, was carried out, in the light of a model based on the concept of a coupling well term - source term. Then, a thermodynamic modelling of ion exchange phenomena in column was established. The formation of the permutation front and the enrichment zones planned was validated by frontal analysis experiments of synthetic fluids (mixtures of Ni(B(OH) 4 ) 2 , LiB(OH) 4 and AgB(OH) 4 in medium B(OH) 3 )), and of real fluid during the putting into service of the device mini-CVCS at the time of Tricastin 2 cold shutdown. New tools are thus proposed, opening the way with an optimised management of demineralizers and a more complete interpretation of the available experience feedback. (author)

  6. Recent results from the MIT in-core experiments on coolant chemistry

    International Nuclear Information System (INIS)

    Harling, O.K.; Kohse, G.E.; Cabello, E.C.; Bernard, J.A.

    1993-01-01

    This paper reports results from an ongoing series of in-core experiments that have been conducted at the 5-MW(thermal) MIT Research Reactor (MITR-II) for optimizing coolant chemistries in light water reactors. Four experiments are in progress, including a pressurized coolant chemistry loop (PCCL), a boiling coolant chemistry loop (BCCL), a facility for the study of irradiation-assisted stress-corrosion cracking, and one for the evaluation of in situ sensors for the monitoring of crack propagation in metal (SENSOR). The first two have now been fully operational for several years. The latter two are scheduled to begin regular operation later this year

  7. Neutrinos: Heavy water detector

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    The proponents of the Sudbury Neutrino Observatory (SNO) received a welcome Christmas present when William Winegard, Canadian Minister for Science and Technology announced the final details of the funding for this project, totalling 48 million Canadian dollars and including contributions from the US and the UK. The SNO experiment will extend significantly the study of solar neutrinos, using some 1,000 tonnes of heavy water to be installed more than two kilometres below ground in a nickel mine at Sudbury, Ontario

  8. Estimation of aluminum and argon activation sources in the HANARO coolant

    International Nuclear Information System (INIS)

    Jun, Byung Jin; Lee, Byung Chul; Kim, Myong Seop

    2010-01-01

    The activation products of aluminum and argon are key radionuclides for operational and environmental radiological safety during the normal operation of open-tank-in-pool type research reactors using aluminum-clad fuels. Their activities measured in the primary coolant and pool surface water of HANARO have been consistent. We estimated their sources from the measured activities and then compared these values with their production rates obtained by a core calculation. For each aluminum activation product, an equivalent aluminum thickness (EAT) in which its production rate is identical to its release rate into the coolant is determined. For the argon activation calculation, the saturated argon concentration in the water at the temperature of the pool surface is assumed. The EATs are 5680, 266 and 1.2 nm, respectively, for Na-24, Mg-27 and Al-28, which are much larger than the flight lengths of the respective recoil nuclides. These values coincide with the water solubility levels and with the half-lives. The EAT for Na-24 is similar to the average oxide layer thickness (OLT) of fuel cladding as well; hence, the majority of them in the oxide layer may be released to the coolant. However, while the average OLT clearly increases with the fuel burn-up during an operation cycle, its effect on the pool-top radiation is not distinguishable. The source of Ar-41 is in good agreement with the calculated reaction rate of Ar-40 dissolved in the coolant

  9. Proposed model for fuel-coolant mixing during a core-melt accident

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1983-01-01

    If complete failure of normal and emergency coolant flow occurs in a light water reactor, fission product decay heat would eventually cause melting of the reactor fuel and cladding. The core melt may then slump into the lower plenum and later into the reactor cavity and contact residual liquid water. A model is proposed to describe the fuel-coolant mixing process upon contact. The model is compared to intermediate scale experiments being conducted at Sandia. The modelling of this mixing process will aid in understanding three important processes: (1) fuel debris sizes upon quenching in water, (2) the hydrogen source term during fuel quench, and (3) the rate of steam production. Additional observations of Sandia data indicate that the steam explosion is affected by this mixing process

  10. [Effect of Recycled Water Irrieation on Heavy Metal Pollution in Irrigation Soil].

    Science.gov (United States)

    Zhou, Yi-qi; Liu, Yun-xia; Fu, Hui-min

    2016-01-15

    With acceleration of urbanization, water shortages will become a serious problem. Usage of reclaimed water for flushing and watering of the green areas will be common in the future. To study the heavy metal contamination of soils after green area irrigation using recycled wastewater from special industries, we selected sewage and laboratory wastewater as water source for integrated oxidation ditch treatment, and the effluent was used as irrigation water of the green area. The irrigation units included broad-leaved forest, bush and lawn. Six samples sites were selected, and 0-20 cm soil of them were collected. Analysis of the heavy metals including Cr, Mn, Ni, Cu, Zn, As, Cd and Pb in the soil showed no significant differences with heavy metals concentration in soil irrigated with tap water. The heavy metals in the soil irrigated with recycled water were mainly enriched in the surface layer, among which the contents of Cr, Ni, Cu, Zn and Pb were below the soil background values of Beijing. A slight pollution of As and Cd was found in the soil irrigated by recycled water, which needs to be noticed.

  11. Investigation of the heavy water distillation system at the RA reactor

    International Nuclear Information System (INIS)

    Zecevic, V.; Badrljica, R.

    1963-01-01

    The heavy water distillation system was tested because this was not done before the reactor start-up. Detailed inspection of the system components showed satisfactory results. Leak testing was done as well as the testing of the instrumentation which enables reliable performance of the system. Performance testing was done with ordinary water and later 2700 l of heavy water from the reactor was purified, decreasing the activity by 45%

  12. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  13. Heavy metal contamination of soil and water in the vicinity of an abandoned e-waste recycling site: implications for dissemination of heavy metals.

    Science.gov (United States)

    Wu, Qihang; Leung, Jonathan Y S; Geng, Xinhua; Chen, Shejun; Huang, Xuexia; Li, Haiyan; Huang, Zhuying; Zhu, Libin; Chen, Jiahao; Lu, Yayin

    2015-02-15

    Illegal e-waste recycling activity has caused heavy metal pollution in many developing countries, including China. In recent years, the Chinese government has strengthened enforcement to impede such activity; however, the heavy metals remaining in the abandoned e-waste recycling site can still pose ecological risk. The present study aimed to investigate the concentrations of heavy metals in soil and water in the vicinity of an abandoned e-waste recycling site in Longtang, South China. Results showed that the surface soil of the former burning and acid-leaching sites was still heavily contaminated with Cd (>0.39 mg kg(-1)) and Cu (>1981 mg kg(-1)), which exceeded their respective guideline levels. The concentration of heavy metals generally decreased with depth in both burning site and paddy field, which is related to the elevated pH and reduced TOM along the depth gradient. The pond water was seriously acidified and contaminated with heavy metals, while the well water was slightly contaminated since heavy metals were mostly retained in the surface soil. The use of pond water for irrigation resulted in considerable heavy metal contamination in the paddy soil. Compared with previous studies, the reduced heavy metal concentrations in the surface soil imply that heavy metals were transported to the other areas, such as pond. Therefore, immediate remediation of the contaminated soil and water is necessary to prevent dissemination of heavy metals and potential ecological disaster. Copyright © 2014 Elsevier B.V. All rights reserved.

  14. Experimental Study on Behavior of Bow-tie Tree Generation by Using Heavy Water

    Science.gov (United States)

    Kumazawa, Takao; Nakagawa, Wataru; Tsurumaru, Hidekazu

    Bow-tie tree (BTT) generated from contaminant, e.g., metal, carbon, amber(over cured resin) or void in insulator is a significant deterioration factor of XLPE power cable. However, essential role of water in generation and progress of BTT is not yet sufficiently cleared. In order to investigate the role of water we paid attention to difference in chemical properties of light water (H2O) and heavy water (D2O), moreover we evaluated influence of isotopic effect due to hydrogen and deuterium on behavior of BTT generation. In accelerated aging test the number of BTT in XLPE sample, in which copper powder of 500ppm was contaminated as BTT cores, dipped in heavy water (D2O:100wt%) decreased to one third compared with light water(D2O:0wt%). Furthermore, the maximum length of BTT decreased with increase in concentration of heavy water. The experimental results show that heavy water exerted a depression effect on generation and progress of BTT. We considered that the depression effect due to hydrogen isotope appeared by inhibiting ionization and elution of BTT cores, because salt-solubility and ionic mobility of heavy water are about 15 to 20% smaller than those of light water. Therefore, the essential role of water seemed to be production and transport of ions in XLPE.

  15. Reactor coolant purification system circulation pumps (CUW pumps)

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

    Coolant purification equipments for BWRs have been improved, and the high pressure purifying system has become the main type. The quantity of purifying treatment also changed to 2% of the flow rate of reactor feed water. As for the circulation pumps, canned motor pumps are adopted recently, and the improvements of reliability and safety are attempted. The impurities carried in by reactor feed water and the corrosion products generated in reactors and auxiliary equipments are activated by neutron irradiation or affect heat transfer adversely, adhering to fuel claddings are core structures. Therefore, a part of reactor coolant is led to the purification equipments, and returned to reactors after the impurities are eliminated perfectly. At the time of starting and stopping reactors, excess reactor water and the contaminated water from reactors are transferred to main condenser hot wells or waste treatment systems. Thus the prescribed water quality is maintained. The operational modes of and the requirements for the CUW pumps, the construction and the features of the canned motor type CUW pumps are explained. Recently, a pump operated for 11 months without any maintenance has been disassembled and inspected, but the wear of bearings has not been observed, and the high reliability of the pump has been proved. (Kako, I.)

  16. Operating performance of the prototype heavy water reactor Fugen

    International Nuclear Information System (INIS)

    1984-01-01

    Since the full scale operation was started in March, 1979, the ATR Fugen power station has been verifying the performance and reliability of the machinery and equipment, uranium-plutonium mixed oxide fuel and so on, and obtaining the technical prospect for putting ATRs in practical use by accumulating operation and maintenance techniques, through about five years of operation. In this report, the operational results of the Fugen power station are described. Fugen is a heavy water-moderated, boiling light water-cooled, pressure tube type reactor with 165 MWe output. As of the end of March, 1984, the total generated electric power was about 4.3 billion kWh, and the operation time was about 27,000 hours. The mean capacity ratio reached 58.8%. During the operation period, troubles including plant shutdown occurred eight times, but generally the performance and reliability of the machinery and equipment have been good. 580 fuels including 284 MOX fuels have been charged, but fuel breaking did not occur at all. The consumption of heavy water and the leak of tritium did not cause problem. The management of the core and fuel, the management of maintenance, the quality control of cooling water and heavy water, radiation control and the management of wastes are reported. (Kako, I.)

  17. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  18. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

    Elliot, A.J.; Stuart, C.R.

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  19. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  20. Lubrication analysis of the journal bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

    2000-01-01

    Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

  1. The Bare Critical Assembly of Natural Uranium and Heavy Water

    Energy Technology Data Exchange (ETDEWEB)

    Popovic, D [Boris Kidric Institute of Nuclear Sciences, Vinca, Belgrade (Yugoslavia)

    1958-07-01

    The first reactor built in Yugoslavia was the bare zero energy heavy water and natural uranium assembly at the Boris Kidric Institute of Nuclear Sciences, Belgrade. The reactor went critical on April 29, 1958. The possession of four tons of natural uranium metal and the temporary availability of seven tons of heavy water encouraged the staff of the Institute to build a critical assembly. A critical assembly was chosen, rather than high flux reactor, because the heavy water was available only temporarily. Besides, a 10 MW, enriched uranium, research reactor is being built at the same Institute and should be ready for operation late this year. It was supposed that the zero energy reactor would provide experience in carrying out critical experiments, operational experience with nuclear reactors, and the possibility for an extensive program in reactor physics. (author)

  2. RETRAN analysis of inter-system LOCA within the primary coolant pump

    International Nuclear Information System (INIS)

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

  3. Neutron disadvantage factors in heavy water and light water reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1966-01-01

    A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

  4. Radiolysis of the VVER-1000 reactor coolant: An experimental study and mathematical modeling

    International Nuclear Information System (INIS)

    Arkhipov, O.P.; Bugaenko, V.L.; Kabakchi, S.A.

    1995-01-01

    Variations in the composition of the coolant for the primary circuit of a VVER-1000 reactor of the Kalinin nuclear power plant upon transition from power-level operation to shutdown was studied experimentally. The data obtained were used for verification of the MORAVA-H2 program developed earlier for simulation of the coolant state in pressurized-water power reactors

  5. Cost effective water treatment program in Heavy Water Plant (Manuguru)

    International Nuclear Information System (INIS)

    Mohapatra, C.; Prasada Rao, G.

    2002-01-01

    Water treatment technology is in a state of continuous evolution. The increasing urgency to conserve water and reduce pollution has in recent years produced an enormous demand for new chemical treatment programs and technologies. Heavy water plant (Manuguru) uses water as raw material (about 3000 m 3 /hr) and its treatment and management has benefited the plant in a significant way. It is a fact that if the water treatment is not proper, it can result in deposit formation and corrosion of metals, which can finally leads to production losses. Therefore, before selecting treatment program, complying w.r.t. quality requirements, safety and pollution aspects cost effectiveness shall be examined. The areas where significant benefits are derived, are raw water treatment using polyelectrolyte instead of inorganic coagulant (alum), change over of regenerant of cation exchangers from hydrochloric acid to sulfuric acid and in-house development of cooling water treatment formulation. The advantages and cost effectiveness of these treatments are discussed in detail. Further these treatments has helped the plant in achieving zero discharge and indirectly increased cost reduction of final product (heavy water); the dosage of 3 ppm of polyelectrolyte can replace 90 ppm alum at turbidity level of 300 NTU of raw water which has resulted in cost saving of Rs. 15-20 lakhs in a year beside other advantages; the change over of regenerant from HCl to H 2 SO 4 will result in cost saving of at least Rs.1.4 crore a year besides other advantages; the change over to proprietary formulation to in-house formulation in cooling water treatment has resulted in a saving about Rs.11 lakhs a year. To achieve the above objectives in a sustainable way the performance results are being monitored. (author)

  6. Method of suppressing the deposition of Co-60 to primary coolant pipeways in a nuclear reactor

    International Nuclear Information System (INIS)

    Hoshi, Michio; Tachikawa, Enzo; Goto, Satoshi; Sagawa, Chiaki; Yonezawa, Chushiro.

    1987-01-01

    Purpose: To suppress the deposition of Co-60 to primary coolant pipeways in a nuclear reactor. Method: To reduce the accumulation of Co-60 by causing chemical species of extremely similar chemical property with soluble Co-60 to be present together in coolants and replacing the deposition of Co-60 to the primary coolant pipeways in a nuclear reactor with that of the coexistent chemical spacies. Ni or Zn is used as the coexistet chemical spacies of similar chemical property with Co-60. The coexistent amount is from 5 to 10 times of the soluble Co-60 in the primary coolants. Ni or Zn solution adjusted with concentration is poured into and mixed with the coolants from a water feed source by using a high pressure constant volume pump. The amount of Co-60 taken into the pipeways caused by corrosion due to high temperature coolant is reduced to about 1/5 as compared with the case of Co-60 alone if 1 ppb of soluble Co-60 is present in water and 5 ppb of soluble Ni or Zn is added and, reduced to 1/12 if the amount of Ni or Zn is 10 ppb. (Kamimura, M.)

  7. Heavy metals concentrations in water bodies around aquamarine ...

    African Journals Online (AJOL)

    Water samples from three streams in the mining area of Eggon Hill were analysed. The Physicochemical values obtained were compared with WHO permissible standards in drinking water. Except for Cu and Zn with levels within permissible limits, other heavy metals determined were found to have levels above the WHO ...

  8. Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

    International Nuclear Information System (INIS)

    Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

    1994-01-01

    Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

  9. Method for separation of water from bituminous shales, etc. [water-free heavy product and water-containing light product

    Energy Technology Data Exchange (ETDEWEB)

    Hellsing, G H

    1908-10-13

    The method is characterized by conducting all the products of distillation, coming from the retorts, into a controllable system of condensation. This system of condensation is so constructed that the products of distillation are cooled to such a temperature that only the water-free heavy distillates are being condensed, and is furthermore so constructed that the other products of distillation, not yet condensed, are being condensed in an ordinary system of coolers. The purpose is to separate the distillates into a water-free heavy product and a water-containing lighter product. The patent includes an additional claim.

  10. Assessment of heavy metals in loose deposits in drinking water distribution system.

    Science.gov (United States)

    Liu, Quanli; Han, Weiqiang; Han, Bingjun; Shu, Min; Shi, Baoyou

    2018-06-09

    Heavy metal accumulation and potential releases from loose deposits in drinking water distribution system (DWDS) can have critical impacts on drinking water safety, but the associated risks have not been sufficiently evaluated. In this work, the potential biological toxicity of heavy metals in loose deposits was calculated based on consensus-based sediment quality guidelines, and the effects of some of the main water quality parameters, such as the pH and bicarbonate and phosphate content, on the release behaviors of pre-accumulated heavy metals were investigated. The results showed that heavy metals (Cu, As, Cr, Pb, and Cd) significantly accumulated in all the samples, but the contents of the heavy metals were multiple magnitudes lower than the Fe and Mn contents. The potential biotoxicity of As and Cu was relatively high, but the biotoxicity of Cd was negligible. The water quality can significantly influence the release of heavy metals from loose deposits. As the pH increased from 7.0 to 9.0, the release of As and Cr obviously increased. The release of As, Cu, Pb, and Cr also accelerated with the addition of phosphate (from 1 to 5 mg/L). In contrast to the trends for the pH and phosphate, variations in the bicarbonate content did not have a significant influence on the release of As and Cr. The release ratios of heavy metals in the samples were very low, and there was not a correlation between the release rate of the heavy metals in the loose deposits and their potential biotoxicity.

  11. Development of a portable heavy-water leak sensor based on laser absorption spectroscopy

    International Nuclear Information System (INIS)

    Lee, Lim; Park, Hyunmin; Kim, Taek-Soo; Kim, Minho; Jeong, Do-Young

    2016-01-01

    Highlights: • We developed a compact and portable laser sensor for a detection of heavy water leakage. • The sensor is wearable and also easy to use to search for the leak point. • It is sensitive enough to find invisible very tiny leaks. - Abstract: A compact and portable leak sensor based on cavity enhanced absorption spectroscopy has been newly developed for a detection of heavy water leakage which may happen in the facilities using heavy water such as pressurized heavy water reactor (PHWR). The developed portable sensor is suitable as an individual instrument for the measuring leak rate and finding the leak location because it is sufficiently compact in size and weight and operated by using an internal battery. In the performance test, the minimum detectable leak rate was estimated as 0.05 g/day from the calibration curve. This new sensor is expected to be a reliable and promising device for the detection of heavy water leakage since it has advantages on real-time monitoring and early detection for nuclear safety.

  12. Sediment, water pollution indicators for heavy metals

    International Nuclear Information System (INIS)

    Cabaleiro, S.; Horn, A.

    2010-01-01

    The complexity of an aquatic system requires consideration of its dynamics: spatial and temporal variations of physical, chemical and biological. Heavy metals have peculiar behavior in the aquatic system and may not be available in the waters, but on sediments.The sub-basin of the Sarandi stream is responsible for the contamination of Pampulha Lake. The Instituto Mineiro das Águas – IGAM - uses tool for monitoring the quality of surface water for developing strategies for conservation, restoration and rational use of water resources. So through the indices: IQA ( Indice de qualidade de águas) Index of water quality, and TC- toxic contamination, reduces conflicts, implements the disciplining of the environmental economy.This study determined the monitoring of sediment and water of Sarandi Stream, so in the samples collected during dry and rainy seasons (2007- 2008) were analyzed heavy metals (Cu, Cd, Cr, Co, Ni, Zn, Pb) and physical-chemical factors (conductivity, solids dissolved, temperature, turbidity). This allowed the determination of Hackanson factors of contamination and Muller Index geoaccumulation, indicating very high contamination in sediments regarding the elements Cr, Cu, and Cd, and high contamination for Pb, Zn, and Mn. The comparison with the indices of water quality- IQA (IGAM - 2006, 2007 and 2008), combined with exploratory data analysis and graphs of correlation between the variables indicated favorable conditions for metals contamination on water and sediment for these metals, besides allowing the identification of its source

  13. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  14. Critical evaluation of heavy water project at Thal (Preprint No. PM-5)

    International Nuclear Information System (INIS)

    Jayakumar, N.S.

    1989-04-01

    The project known as Thal Ammonia Extension was a heavy water project successfully completed by Rashtriya Chemicals and Fertilizers (RCF) Ltd. The project consisted of erecting a heavy water plant of 110 tons/year capacity at Thal. The process Know-how and engineering of the plant was supplied by the Heavy Water Projects Division of the Department of Atomic Energy. Salient features of the project, management features which resulted in fast completion of erection, bottlenecks faced and engineering innovations adopted for efficient operation of the plant are described. Some modifications which can lead to smoother operation are listed. (M.G.B.)

  15. Correlation of cylinder-head temperatures and coolant heat rejections of a multicylinder, liquid-cooled engine of 1710-cubic-inch displacement

    Science.gov (United States)

    Lundin, Bruce T; Povolny, John H; Chelko, Louis J

    1949-01-01

    Data obtained from an extensive investigation of the cooling characteristics of four multicylinder, liquid-cooled engines have been analyzed and a correlation of both the cylinder-head temperatures and the coolant heat rejections with the primary engine and coolant variables was obtained. The method of correlation was previously developed by the NACA from an analysis of the cooling processes involved in a liquid-cooled-engine cylinder and is based on the theory of nonboiling, forced-convection heat transfer. The data correlated included engine power outputs from 275 to 1860 brake horsepower; coolant flows from 50 to 320 gallons per minute; coolants varying in composition from 100 percent water to 97 percent ethylene glycol and 3 percent water; and ranges of engine speed, manifold pressure, carburetor-air temperature, fuel-air ratio, exhaust-gas pressure, ignition timing, and coolant temperature. The effect on engine cooling of scale formation on the coolant passages of the engine and of boiling of the coolant under various operating conditions is also discussed.

  16. Reduction of corrosion products in water coolant - basic way of increase in efficiency and improvement of ecological safety of NPU

    International Nuclear Information System (INIS)

    Prozorov, V.V.

    2004-01-01

    Corrosion of oxidated steel in water with additives of inhibitors or oxygen was considered. It is shown that preliminary oxidation of steel makes possible declining concentration of inhibitors or oxygen. Experiments demonstrate possibilities of the neutral-oxygen water regime for supply of the effective protection. Corrosion resistance of steel may be increased in many times through correct aqua-chemical regimes. Also concentration of corrosion products may be decreased in many times in coolant and their activation in neutron flux of nuclear reactor, amount of radioisotopes [ru

  17. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

  18. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  19. On natural circulation in High Temperature Gas-Cooled Reactors and pebble bed reactors for different flow regimes and various coolant gases

    International Nuclear Information System (INIS)

    Melesed'Hospital, G.

    1983-01-01

    The use of CO 2 or N 2 (heavy gas) instead of helium during natural circulation leads to improved performance in both High Temperature Gas-Cooled Reactors (HTGR) and in Pebble Bed Reactors (PBR). For instance, the coolant temperature rise corresponding to a coolant pressure level and a rate of afterheat removal could be only 18% with CO 2 as compared to He, for laminar flow in HTGR; this value would be 40% in PBR. There is less difference between HTGR and PBR for turbulent flows; CO 2 is found to be always better than N 2 . These types of results derived from relationships between coolant properties, coolant flow, temperature rise, pressure, afterheat levels and core geometry, are obtained for HTGR and PBR for various flow regimes, both within the core and in the primary loop

  20. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  1. Phytoremediation of heavy metal-contaminated water and sediment by eleocharis acicularis

    Energy Technology Data Exchange (ETDEWEB)

    Sakakibara, Masayuki; Ha, Nguyen Thi Hoang [Graduate School of Science and Engineering, Ehime University, Matsuyama (Japan); Ohmori, Yuko [Graduate School of Science and Engineering, Ehime University, Matsuyama (Japan); Taisei Kiso Sekkei Co., Ltd., Tokyo (Japan); Sano, Sakae [Faculty of Education, Ehime University, Matsuyama (Japan); Sera, Koichiro [Cyclotron Center, Iwate Medical University, Takizawa-mura (Japan)

    2011-08-15

    Phytoremediation is an environmental remediation technique that takes advantage of plant physiology and metabolism. The unique property of heavy metal hyperaccumulation by the macrophyte Eleocharis acicularis is of great significance in the phytoremediation of water and sediments contaminated by heavy metals at mine sites. In this study, a field cultivation experiment was performed to examine the applicability of E. acicularis to the remediation of water contaminated by heavy metals. The highest concentrations of heavy metals in the shoots of E. acicularis were 20 200 mg Cu/kg, 14 200 mg Zn/kg, 1740 mg As/kg, 894 mg Pb/kg, and 239 mg Cd/kg. The concentrations of Cu, Zn, As, Cd, and Pb in the shoots correlate with their concentrations in the soil in a log-linear fashion. The bioconcentration factor for these elements decreases log-linearly with increasing concentration in the soil. The results indicate the ability of E. acicularis to hyperaccumulate Cu, Zn, As, and Cd under natural conditions, making it a good candidate species for the phytoremediation of water contaminated by heavy metals. (Copyright copyright 2011 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. Assessment of heavy metals concentration in drinking water ...

    African Journals Online (AJOL)

    The concentration of all the metals were considerably found to be below the limit permitted by WHO's drinking water guidelines (WHO 2005). Findings suggest that continues water quality monitoring should be carried out to check the concentration levels of heavy metals in that area, to prevent them from been above the limit ...

  3. Analysis of hydrogen sulfide releases in heavy water production facilities

    International Nuclear Information System (INIS)

    Croitoru, Cornelia; Dumitrescu, Maria; Preda, Irina; Lazar, Roxana

    1996-01-01

    Safety analyses conducted at ICIS concern primarily the heavy water production installations. The quantitative risk assessment needs the frequency calculation of accident sequences and consequences. In heavy water plants which obtain primary isotopic concentration of water by H 2 O - H 2 S exchange, large amounts of hydrogen sulfide which is a toxic, inflammable and explosive gas, are circulated. The first stage in calculating the consequences consists in potential analysis of H 2 S release. This work presents a study of this types of releases for pilot installations of the heavy water production at ICIS (Plant 'G' at Rm. Valcea). The installations which contain and maneuver large quantities of H 2 S and the mathematical models for different types of releases are presented. The accidents analyzed are: catastrophic column, container, spy-hole failures or gas-duct rupture and wall cracks in the installation. The main results are given as tables while the time variations of the flow rate and quantities of H 2 O released by stack disposal are plotted

  4. Winter Maintenance Wash-Water Heavy Metal Removal Pilot Scale Evaluation

    Directory of Open Access Journals (Sweden)

    Christopher M. Miller

    2016-01-01

    Full Text Available To encourage sustainable engineering practices, departments of transportation are interested in reusing winter maintenance truck wash water as part of their brine production and future road application. Traffic-related metals in the wash water, however, could limit this option. The objective of this work was to conduct a pilot scale evaluation of heavy metal (copper, zinc, iron, and lead removal in a filtration unit (maximum flow rate of 45 L/minute containing proprietary (MAR Systems Sorbster® media. Three different trials were conducted and approximately 10,000 L of wash water collected from a winter maintenance facility in Ohio was treated with the pilot unit. Lab studies were also performed on six wash-water samples from multiple facilities to assess particle size removal and estimate settling time as a potential removal mechanism during wash-water storage. Pilot unit total metal removal efficiencies were 79%, 77%, 63%, and 94% for copper, zinc, iron, and lead, respectively. Particle settling calculation estimates for copper and zinc show that 10 hours in storage can also effectively reduce heavy metal concentrations in winter maintenance wash water in excess of 70%. These pilot scale results show promise for reducing heavy metal concentrations to an acceptable level for reuse.

  5. Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

    International Nuclear Information System (INIS)

    Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

    2001-01-01

    In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

  6. Parametric thermal analysis of 75 MHz heavy ion RFQ

    International Nuclear Information System (INIS)

    Mishra, N.K.; Mehrotra, N.; Verma, V.; Gupta, A.K.; Bhagwat, P.V.

    2015-01-01

    An ECR based Heavy Ion Accelerator comprising of a superconducting Electron Cyclotron Resonance (ECR) Ion Source, normal conducting RFQ (Radio Frequency Quadrupole) and superconducting Niobium resonators is being developed at BARC under XII plan. A state-of-the-art 18 GHz superconducting ECR ion source (PK-ISIS) jointly configured with Pantechnik, France is operational at Van-de-Graaff, BARC. The electromagnetic design of the improved version of 75 MHz heavy ion RFQ has been reported earlier. The previous thermal study of 51 cm RFQ model showed large temperature variation axially along the vane tip. A new coolant flow scheme has been worked out to optimize the axial temperature gradient. In this paper the thermal analysis including parametric study of coolant flow rates and inlet temperature variation will be presented. (author)

  7. Feasibility study and economic analysis on thorium utilization in heavy water reactors

    International Nuclear Information System (INIS)

    1978-07-01

    Even though natural uranium is a more easily usable fuel in heavy water reactors, thorium fuel cycles have also been considered owing to certain attractive features of the thorium fuel cycle in heavy water reactors. The relatively higher fission neutron yield per thermal neutron absorption in 233 U combined with the very low neutron absorption cross section of heavy water make it possible to achieve breeding in a heavy water reactor operating on Th- 233 U fuel cycle. Even if the breeding ratio is very low, once a self-sustaining cycle is achieved, thereafter dependence on uranium can be completely eliminated. Thus, with a self-sustaining Th- 233 U fuel cycle in heavy water reactors, a given quantity of natural uranium will be capable of supporting a much larger installed generating capacity to significantly longer period of time. However, since thorium does not contain any fissile isotope, fissile material has to be added at the beginning. Concentrated fissile material is considerably more expensive than the 235 U contained in natural uranium. This makes the fuel cycle cost higher with thorium fuel cycle, at least during the initial stages. The situation is made worse by the fact that, because of its higher thermal neutron absorption cross section, thorium requires a higher concentration of fissile material than 238 U. Nevertheless, because of the superior nuclear characteristics of 233 U, once uranium becomes more expensive, thorium fuel cycle in heavy water reactors may become economically acceptable. Furthermore, the energy that can be made available from a given quantity of uranium is considerably increased with a self-sustaining thorium fuel cycle

  8. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  9. RA Reactor operation and maintenance (I-IX), part VIII, Task 3.08/05, Decontamination of the reactor

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    Permanent increase of radiation in the heavy water system was noticed during first three year of the RA reactor operation, even when the reactor was shutdown. It was found that there was no failure of the fuel element cladding. Radioactive cobalt was found in the heavy water which was rather strange. During repair of the heavy water system, it has been found that stellite was used for coating the heavy water pumps. Since stellite is a cobalt alloy, this could have been the source of radioactive cobalt in the heavy water. The stellite coating was damaged due to friction and particle of cobalt appeared in the coolant, they were activated since they were in the core. decontamination of the heavy water and the heavy water coolant loop was a must . Beside the detailed report on the contamination and decontamination of the heavy water system this volume includes 14 annexes describing the investigation of the event and the whole procedure of decontamination

  10. RA Reactor operation and maintenance (I-IX), part VIII, Task 3.08/05, Decontamination of the reactor; Pogon i odrzavanje reaktora (I-IX), VIII Deo, Zadatak 3.08/05 Dekontaminacija reaktora

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-12-15

    Permanent increase of radiation in the heavy water system was noticed during first three year of the RA reactor operation, even when the reactor was shutdown. It was found that there was no failure of the fuel element cladding. Radioactive cobalt was found in the heavy water which was rather strange. During repair of the heavy water system, it has been found that stellite was used for coating the heavy water pumps. Since stellite is a cobalt alloy, this could have been the source of radioactive cobalt in the heavy water. The stellite coating was damaged due to friction and particle of cobalt appeared in the coolant, they were activated since they were in the core. decontamination of the heavy water and the heavy water coolant loop was a must . Beside the detailed report on the contamination and decontamination of the heavy water system this volume includes 14 annexes describing the investigation of the event and the whole procedure of decontamination.

  11. Coolant Design System for Liquid Propellant Aerospike Engines

    Science.gov (United States)

    McConnell, Miranda; Branam, Richard

    2015-11-01

    Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

  12. Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

    Science.gov (United States)

    Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  13. Method of extracting tritium from heavy water

    International Nuclear Information System (INIS)

    Tsuchiya, Hiroyuki; Kikuchi, Makoto; Asakura, Yamato; Yusa, Hideo.

    1979-01-01

    Purpose: To extract tritium in heavy water by combining isotope exchange reaction with liquefaction distillation to increase the concentration of recovered tritium, thereby reducing the quantity of radioactive wastes recovered. Constitution: Heavy water containing tritium from a reactor is introduced into a tritium separator through a conduit pipe. On the other hand, a D 2 gas is introduced through the conduit pipe in the lower part of a tritium separator to transfer tritium into D 2 gas by isotope exchange. The D 2 gas containing DT is introduced into a liquefaction distillation tower together with an outlet gas of a converter supplied through a pipeline. The converter is filled with net-like metals of platinum group such as Pt, Ni, Pd and the like, and the D 2 gas affluent in DT, extracted from the distillation tower is converted into D 2 and T 2 . The gas which has been introduced into the liquefaction distillation tower is liquefied. The D 2 gas of low boiling point components reaches the tower top, and the T 2 gas of high boiling point components is concentrated at the tower bottom, and is rendered into tritium water in a recoupler and stored in a water storage apparatus. (Yoshino, Y.)

  14. The projects for heavy water production of the Argentine National Atomic Energy Commission

    International Nuclear Information System (INIS)

    Garcia Bourg, J.M.; Garcia, E.E.

    1982-01-01

    The bases and scope of the projects for heavy water production that are being currently developed by the Argentine National Atomic Energy Commission (CNEA) are described. As an introduction, the following points are presented: a) the fundamentals of heavy water utilization in a nuclear reactor, with a mention of its properties and uses, b) a review of the physicochemical bases of the principal methods for heavy water production: chemical exchange (monothermal and bithermal processes), distillation and electrolysis, with tables summarizing the fundamental characteristics of the first two ones, and an evaluation of the different production methods from the viewpoint of their application in an industrial scale; and c) a synthetic information, in the form of tables, about the world's heavy water production. The subject of heavy water production in Argentina is treated in the principal section, describing the scope, location, main characteristics and chemical processes corresponding to the projects being developed by CNEA, which currently are the installation of an Industrial Plant in Arroyito (Province of Neuquen), purchased on a turnkey basis and using the NH 3 /H 2 isotopic exchange method; the installation of an Experimental Plant in Atucha (Province of Buenos Aires), for the development of the domestic technology of heavy-water production by the SH 2 /H 2 O isotopic exchange method, and the development of the engineering of an industrial plant (''Module 80''), based on the Experimental Plant's technology. (M.E.L.) [es

  15. Safety Evaluation of Osun River Water Containing Heavy Metals and ...

    African Journals Online (AJOL)

    Summary: This study evaluated the pH, heavy metals and volatile organic compounds (VOCs) in Osun river water. It also evaluated its safety in rats. Heavy metals were determined by atomic absorption spectrophotometry (AAS) while VOCs were determined by gas chromatography coupled with flame ionization detector ...

  16. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

  17. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  18. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  19. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

    Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

    1969-01-01

    Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

  20. Design criteria of primary coolant chemistry in SMART-P

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

  1. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  2. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  3. Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition)

    International Nuclear Information System (INIS)

    2010-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

  4. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  5. Assessment of Heavy Metals in the Water of Sahastradhara Hill Stream at Dehradun, India

    Directory of Open Access Journals (Sweden)

    Pawan Kumar Bharti

    2014-09-01

    Full Text Available A study on heavy metals assessment in the water of Sahastradhara hill-stream was conducted with different five sites at significant differences. The present paper deals with the water quality status of Sahastradhara stream by the assessment of heavy metals. Heavy Metals were found in fluctuated trend from first upstream to last downstream. The values of almost all Heavy Metals were found in increasing manner especially after the fourth sampling site. After the third sampling station, a solid waste dumping site was found. So, there may be a relation between heavy metals in stream water and solid waste dumping site. Concentrations of all Heavy Metals at fourth and fifth sampling site were found very high. DOI: http://dx.doi.org/10.3126/ije.v3i3.11076 International Journal of Environment Vol.3(3 2014: 164-172

  6. Electrolytic separation factors for oxygen isotopes in light and heavy water solutions

    International Nuclear Information System (INIS)

    Gulens, J.; Olmstead, W.J.; Longhurst, T.H.; Gale, K.L.; Rolston, J.H.

    1987-01-01

    The electrolytic separation factor, α, has been measured for /sup 17/O and /sup 18/O at Pt and Ni anodes in both light and heavy water solutions of 6M KOH as a function of current density. For oxygen-17, isotopic separation effects were not observed, within the experimental uncertainty of +-2%, under all conditions studied. For oxygen-18, there is a small difference of 2% in α values between Pt and Ni in both light and heavy water solutions, but there is no significant difference in α values between light and heavy water solutions. In light waters solutions, the separation factor at Pt is small, α(/sup 18/O) ≤ 1.02 for i ≥ 0.1 A/cm/sub 2/. This value agrees reasonably well with theoretical estimates

  7. Pulse radiolysis studies of liquid heavy water at temperatures up to 250 degrees C

    International Nuclear Information System (INIS)

    Stuart, C.R.; Ouellette, D.C.; Elliot, A.J.

    2002-09-01

    This report documents the rate constants and associated activation energies for the reactions of the primary radical species, e aq - , ·OD and ·D, which are formed during the radiolysis of heavy water within the temperature range 20 to 250 o C. These heavy-water data have been compared with the corresponding information for light water. These kinetic data form part of the database that is required to model the aqueous radiation chemistry that occurs within the core of the heavy water cooled and moderated CANDU reactor. (author)

  8. Pulse radiolysis studies of liquid heavy water at temperatures up to 250 degrees C

    Energy Technology Data Exchange (ETDEWEB)

    Stuart, C.R.; Ouellette, D.C.; Elliot, A.J

    2002-09-01

    This report documents the rate constants and associated activation energies for the reactions of the primary radical species, e{sub aq}{sup -}, {center_dot}OD and {center_dot}D, which are formed during the radiolysis of heavy water within the temperature range 20 to 250 {sup o}C. These heavy-water data have been compared with the corresponding information for light water. These kinetic data form part of the database that is required to model the aqueous radiation chemistry that occurs within the core of the heavy water cooled and moderated CANDU reactor. (author)

  9. The Battle for Heavy Water Three physicists' heroic exploits

    CERN Multimedia

    2002-01-01

    Up until the end of the 1970s you could still catch a glimpse of his massive silhouette in the corridors of CERN. Lew Kowarksi, one of the pioneers of the Laboratory, was not only a great physicist; he was also a genuine hero of World War II. In 1940, along with Frédéric Joliot and Hans von Halban, Lew Kowarski managed to get the entire world supply of heavy water away to safety from the Nazis after a fantastic escape from occupied France. At the end of the war, the three physicists played themselves in a film about their adventures entitled 'la Bataille de l'eau lourde'. This film, which has been loaned to us by the French National Film Library, will be shown at CERN for the first time next Thursday. At the beginning of the war, heavy water (D20, two atoms of deuterium and one oxygen atom) was of strategic importance. In 1939 Frédéric Joliot, aided by Hans von Halban and Lew Kowarski, demonstrated the nuclear chain reaction and the moderator role that heavy water plays in it. A few weeks before the inv...

  10. The importance of heavy water in nuclear technology

    International Nuclear Information System (INIS)

    Gharib, A.G.

    2004-01-01

    Due to similarities of chemical and almost physical properties in H 2 O and D 2 O but differences in nuclear and particle peculiarities provide valuable application for D 2 O. To sustain a controlled chain reaction, the energy of neutrons produced by fission must be reduced through collisions with other nuclei, a process called moderation. A good moderator has a mass close to that of the neutron to maximize energy loss per collision and a very small neutron capture cross section to minimize unwanted nuclear reactions. Deuterium is far the best moderator, more than 80 times better than hydrogen and 30 times better than 12 C ir 18 O. Heavy water is almost as good as deuterium and has the distinct advantage of being a nonflammable liquid. Heavy water is also an excellent neutron reflector, and thus decreases the number of neutrons that escape the reactor core without participating in fission reactions. For this reason a feasibility study and subsequently a technical survey was carried out on engineering of a pilot scale plant. As the result of this studies, the know-how of heavy water production on basis of selected method including dual temperature isotopic exchange and distillation techniques developed. Subsequently the primary and almost detail engineering documents prepared on best knowledge of our own engineers without external contribution

  11. Direct potentiometric control of chloride-ion content in water coolant of nuclear reactors

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Vilkov, N.Ya.; Krasnoperov, V.M.; Epimakhova, L.V.

    1979-01-01

    The work of automatic chloride measuring device designed for continuous determination of chloride-ion concentration in water coolants of nuclear power plants is investigated. A series of experiments have been performed to investigate a device with sensitive element in the form of potentiometric cell with two flowing porous metal silver electrodes (PSE), placed in series. A calibration circuit of chloride measuring devices and PSE is described. A comparison is made between the results obtained by means of automatic chloride measuring device and results of manual control of samples. A conclusion is drawn that automatic chloride measuring devices meet the requirements of nuclear power plants for methods and instruments of control of chloride-ions microconcentration. The development and implantation of automatic chloride-ion analizers will make the analytical control on nuclear power plants easier and make it possible to obtain more reliable information

  12. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

  13. Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

    International Nuclear Information System (INIS)

    Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

    2013-01-01

    The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

  14. Air water loop - an experimental facility to study thermal hydraulics of AHWR steam drum

    International Nuclear Information System (INIS)

    Bagul, R.K.; Pilkhwal, D.S.; Jain, V.; Vijayan, P.K.

    2014-05-01

    In the proposed Indian Advanced Heavy Water Reactor (AHWR) the coolant recirculation in the primary system is achieved by two-phase natural circulation. The two-phase steam-water mixture from the reactor core is separated in steam drum by gravity. Gravity separation of phases may lead to undesirable phenomena - carryover and carryunder. Carryover is the entrainment of liquid droplets in the vapor phase.Carryover needs to be minimized to avoid erosion corrosion of turbine blades. Carryunder is the entrainment of vapor bubbles with liquid flowing back to reactor core. Significant carryunder may in turn lead to reduced flow resulting in reduced CHF margin and stability in the coolant channel. An Air-Water Loop (AWL) has been designed to carry out the experiments relevant to AHWR steam drum. The design features and scaling philosophy is described in this report. (author)

  15. Effects of molten material temperatures and coolant temperatures on vapor explosion

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

    2007-01-01

    An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

  16. Special operations in the heavy water system, III-2; III-2 Posebne operacije u sistemu teske vode

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1989-07-01

    Special operations in the heavy water system described in this chapter are as follows: treatment, drainage and pouring of heavy water, emptying of the heavy water system, cleaning and vacuuming of the heavy water system. [Serbo-Croat] Posebne operacije u sistemu teske vode opisane u ovom poglavlju su: tretiranje, dreniranje i razlivanje teske vode, praznjenje sistema teske vode, 'ispiranje' i vakumiranje sistema teske vode.

  17. Improvements on heavy water separation technology by isotopic water-hydrogen sulfide exchange

    International Nuclear Information System (INIS)

    Peculea, M.

    1987-01-01

    A series of possible variance is presented for the heavy water separation technology by isotopic H 2 O-H 2 S exchange at dual temperatures. The critical study of these variants, which are considered as characteristic quantities for the isotopes transport (production) and the extraction level is related to a dual temperature plant fed by liquid and cold column, which is the up-to-date technology employed in all heavy water production plants as variants of following plants are studied: dual temperature plant with double feeding; dual-temperature plant with equilibrium column (booster); dual-temperature-dual-pressure plant. Attention is paid to the variant with equilibration column (booster), executed and tested at the State Committee for Nuclear Energy and to the dual-temperature-dual pressure plant which presents the highest efficiency. (author)

  18. Surface Water Modeling Using an EPA Computer Code for Tritiated Waste Water Discharge from the heavy Water Facility

    International Nuclear Information System (INIS)

    Chen, K.F.

    1998-06-01

    Tritium releases from the D-Area Heavy Water Facilities to the Savannah River have been analyzed. The U.S. EPA WASP5 computer code was used to simulate surface water transport for tritium releases from the D-Area Drum Wash, Rework, and DW facilities. The WASP5 model was qualified with the 1993 tritium measurements at U.S. Highway 301. At the maximum tritiated waste water concentrations, the calculated tritium concentration in the Savannah River at U.S. Highway 301 due to concurrent releases from D-Area Heavy Water Facilities varies from 5.9 to 18.0 pCi/ml as a function of the operation conditions of these facilities. The calculated concentration becomes the lowest when the batch releases method for the Drum Wash Waste Tanks is adopted

  19. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

    1986-01-01

    Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

  20. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

    The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

  1. Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Konovalenko, Alexander, E-mail: kono@kth.se; Karbojian, Aram, E-mail: karbojan@kth.se

    2017-04-01

    Highlights: • Steam explosion in stratified melt-coolant configuration is studied experimentally. • Different binary oxidic melt simulant materials were used. • Five spontaneous steam explosions were observed. • Instability of melt-coolant interface and formation of premixing layer was observed. • Explosion strength is influenced by melt superheat and water subcooling. - Abstract: Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.

  2. Commissioning performance activities of Heavy Water Plant (Hazira) (Paper No. 1.4)

    International Nuclear Information System (INIS)

    Anon.

    1992-01-01

    Heavy Water Plant, Hazira is the fourth in the line of plants based on monothermal NH 3 -H 2 exchange process. The experience gained during operation of other heavy water plants is reflected in the construction, commissioning and operation of HWP, Hazira. This paper aims at outlining the strategy adopted for both commissioning and operation. (author)

  3. Programmable logic controllers in Heavy Water Project, Manuguru (Paper No. 3.4)

    International Nuclear Information System (INIS)

    Gupta, S.C.; Bhaskar, R.; Maiti, A.; Venkatesu, G.; Satish, P.; Goel, R.K.

    1992-01-01

    Enhancement to plant operational flexibility has been achieved in Heavy Water Project, Manuguru by installing programmable logic controllers for its control equipment. The earlier sulfide based Heavy Water Plant, Kota is using relay logic and diode based program-matrix for binary controls. Performance improvement and advantages of PLC and experience in its operation are described. (author). 3 refs

  4. Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, B.R.

    1986-01-01

    In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

  5. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  6. Fluoride content in water in and around heavy water plant Manuguru colony

    International Nuclear Information System (INIS)

    Mohapatra, C.; Dubey, S.K.; Reddy, A.R.; Ravi Kumar, T.S.P.; Selvaraj, S.

    1996-01-01

    Fluoride concentration in water used for human consumption has significant importance with respect to its toxic effects. Hence there was a need for analysing fluoride concentration in drinking water primarily used at Heavy water Plant, Manuguru (HWP (M)) colony and its nearby villages. We found that at HWP (M) colony there is not much variation in the fluoride concentration. However, nearby villages are having wide variation from 0.79 to 5.1 ppm. (author). 5 refs., 1 tab

  7. Assets optimization at Heavy Water Plants

    International Nuclear Information System (INIS)

    Hiremath, S.C.

    2006-01-01

    In the world where the fittest can only survive, manufacturing and production enterprises are under intense pressure to achieve maximum efficiency in each and every field related to the ultimate production of plant. The winners will be those that use their assets, i.e men, material, machinery and money most effectively. The objective is to optimize the utilization of all plant assets-from entire process lines to individual pressure vessels, piping, process machinery, and vital machine components. Assets of Heavy Water Plants mainly consist of Civil Structures, Equipment and Systems (Mechanical, Electrical) and Resources like Water, Energy and Environment

  8. Features in ammonia plant for maximising heavy water production (Paper No. 2.10)

    International Nuclear Information System (INIS)

    Tangri, N.N.; Singh, R.J.; Mukherjee, P.K.; Mishra, B.N.

    1992-01-01

    Whenever an ammonia plant is linked with heavy water production, a system should be foreseen in the design stage itself for total conservation of D 2 in synthesis gas and zero D 2 loss. The process should ensure recycle of D 2 rich condensate within the front end. This alone would be the single most important factor for improving heavy water production rate. The synthesis loop pressure should be chosen keeping in view the interest of heavy water plant (HWP). With vast experience in engineering NH 3 and HWP plants, it is possible to integrate HWP requirements at the design stage itself. (author)

  9. Urban water pollution by heavy metals and health implication in ...

    African Journals Online (AJOL)

    Studies of common heavy metals were conducted at Onitsha, Anambra State, the most urbanized city in Southeastern Nigeria. It was discovered that both surface and subsurface water were heavily polluted. Seven (7) heavy metals namely: arsenic (As+2), cadmium (Cd+2), lead (Pb+2), mercury (Hg+2), zinc (Zn+2), copper ...

  10. Application of heat-resistant non invasive acoustic transducers for coolant control in the NPP pipelines

    International Nuclear Information System (INIS)

    Melnikov, V.; Nigmatulin, B.

    1997-01-01

    The use of ultrasonic waves enables remote testing of the coolant flow, detection of solid and gaseous occlusions and measuring of the water velocity and level. Analysis of the acoustic noise makes it possible to detect coolant leaks and diagnose the state and operation of the rotating mechanisms and bearings. Results are given of the research in the development of highly reliable waveguide-type non-invasive acoustic transducers with a long service life. Examples are given of the use of transducers in various fields of nuclear technology: detection of gas in coolant, indication of the coolant level, control of pipe filling and drainage, measurement of liquid film velocity at the pipe inner surface. (M.D.)

  11. Fuel performance in NPD while operating with two-phase coolant

    International Nuclear Information System (INIS)

    Bain, A.S.

    1978-03-01

    The NPD reactor operated as a boiling heavy water reactor from October 27, 1968 to April 18, 1971. At 25 MWe the steam quality at the steam generator inlet was 13 wt%, and fuel channel outlet steam qualities ranged from 2 to 22 wt%. During this period ammonia was used for oxygen suppression and pH control. At equilibrium the coolant had 7 mg NH 3 /kg D 2 O, 60 ml D 2 /kg D 2 O and 20 ml N 2 /kg D 2 O. The performance of the fuel was excellent during the time that NPD operated in the boiling mode. No indications were observed of dimensional changes, inter-element fretting, fuel/sheath interaction, excessive oxidation, excessive deuterium concentrations, or unusual migration of hydrogen and deuterium to the cooler end plugs. One element defected; although the defect mechanism could not be identified at the time, we now believe the defect was associated with faulty bar stock for end plugs. The behaviour of the defective element was similar to that for other defective elements in CANDU reactors. No problems were encountered in removing the defected bundle from the reactor. (author)

  12. Heavy-water extraction from non-electrolytic hydrogen streams

    International Nuclear Information System (INIS)

    LeRoy, R.L.; Hammerli, M.; Butler, J.P.

    1981-01-01

    Heavy water may be produced from non-electrolytic hydrogen streams using a combined electrolysis and catalytic exchange process. The method comprises contacting feed water in a catalyst column with hydrogen gas originating partly from a non-electrolytic hydrogen stream and partly from an electrolytic hydrogen stream, so as to enrich the feed water with the deuterium extracted from both the non-electrolytic and electrolytic hydrogen gas, and passing the deuterium water to an electrolyser wherein the electrolytic hydrogen gas is generated and then fed through the catalyst column. (L.L.)

  13. The 10B(n,α)7Li reaction in PWR coolants: calculations of the effect on coolant pH and on decreases in 10B isotopic fractions

    International Nuclear Information System (INIS)

    Polley, M.V.

    1988-07-01

    Boron is used as a chemical shim in PWRs for reactivity control and is added in the form of boric acid to the primary coolant. The 10 B(n,α) 7 Li reaction leads to a continuous increase in 7 Li in the primary coolant and to a continuous decrease in 10 B the isotope of boron responsible for control of reactivity. The rate of increase in coolant pH due to 7 Li production is calculated for the Sizewell 'B' PWR to enable judgements to be made on the frequency of sampling and removal of lithium required to maintain the pH of the primary coolant within the desired limits. Calculations are contrasted for the cases of natural boron and 100% 10 B chemical shims, for both a normal cycle and an extended 18 month cycle. Calculations of 10 B depletion over 30 years of operation as a function of the quantity of boron discharged to waste are also presented. 10 B isotopic fractions are calculated for the reactor coolant (RC), boric acid tanks (BATs) and refuelling water storage tank (RWST) assuming rapid mixing of BAT and RC boron for tritium control and other reasons. Such predictions enable assessments of the reactor physics implications of 10 B consumption to be made. (author)

  14. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during the 1970's and early 1980's raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  15. Reactor coolant pump seal issues and their applicability to new reactor designs

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at U.S. operating plants during the 1970's and early 1980's raised concerns from the U.S. Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants

  16. Environmental health scoping study at Bruce Heavy Water Plant

    International Nuclear Information System (INIS)

    Prior, M.; Mostrom, M.; Coppock, R.; Florence, Z.

    1995-10-01

    There are concerns that hydrogen sulfide released from the Heavy Water Plant near Kincardine, Ontario may be the cause of the mortalities and morbidities observed in a nearby flock of sheep. The Philosopher's Wool sheep farm is about four kilometres south-southeast of the Bruce Heavy Water Plant. Ontario Hydro, the owner and operator of the Bruce Heavy Water Plant, claims that hydrogen sulphide emissions from the Bruce Heavy Water Plant are within regulatory limits and well below levels that cause harm. Accordingly, the Atomic Energy Control Board commissioned the Alberta Environmental Centre, Alberta Department of Environmental Protection, to develop a scoping study for this environmental health issue. The first objective was to describe a field investigation model to define clearly the environmental health and operation of the sheep farm. The second objective was to describe possible exposure patterns and develop a holistic environmental pathway model. If appropriate, the third study objective was to describe animal models of the actual situation to elucidate specific aspects of the environmental health concerns. It was not the objective of this report to provide a definitive answer to the present environmental health issue. Ontario Hydro provided data to the Alberta Environmental Centre, as di the sheep farmer, the attending veterinarian, the University of Guelph study team, and the Atomic Energy Control Board. A six-tiered strategy of sequential evaluations of the ovine health problem is based on the multiple-response paradigm. It assumes the observed ovine health results are the result of multiple effector events. Each tier constitutes a separate, but inter-related, study. Sequential evaluation and feedback of each tier allow sound scientific judgements and efficient use of resources. (author). 59 refs., 11 tabs., 22 figs

  17. An on-line tritium-in-water monitor

    International Nuclear Information System (INIS)

    Singh, A.N.; Ratnakaran, M.; Vohra, K.G.

    1985-01-01

    The paper describes the development and operation of a continuous on-line tritium-in-water monitor for the detection of heavy water leaks into the secondary coolant light water of a heavy water power reactor. The heart of the instrument is its plastic scintillator sponge detector, made from 5 μm thick plastic scintillator films. The sponge weighs only about 1 g and is in the form of disc of 48 mm diameter and 8 mm thickness. The total surface area of the films is about 3000 cm 2 . In the coincidence mode of counting, the detector gives 1000 cps for the passage of 3.7 x 10 4 Bq/cm 3 (1 μCi/cm 3 ) of tritiated water. The background in 6 cm thick lead shielding in the laboratory is 0.2 cps, and inside the reactor building it is below 1 cps. The monitor presently scans 18 sample lines in sequence for 5 min each and gives a printout for the activity in each line. (orig.)

  18. An on-line tritium-in-water monitor

    Science.gov (United States)

    Singh, A. N.; Ratnakaran, M.; Vohra, K. G.

    1985-05-01

    The paper describes the development and operation of a continuous on-line tritium-in-water monitor for the detection of heavy water leaks into the secondary coolant light water of a heavy water power reactor. The heart of the instrument is its plastic scintillator sponge detector, made from 5 μm thick plastic scintillator films. The sponge weighs only about 1 g and is in the form of disc of 48 mm diameter and 8 mm thickness. The total surface area of the films is about 3000 cm 2. In the coincidence mode of counting, the detector gives 1000 cps for the passage of 3.7 × 10 4 Bq/cm 3 (1 μCi/cm 3) of tritiated water. The background in 6 cm thick lead shielding in the laboratory is 0.2 cps, and inside the reactor building it is below 1 cps. The monitor presently scans 18 sample lines in sequence for 5 min each and gives a printout for the activity in each line.

  19. Solubilities of boric acid in heavy water

    International Nuclear Information System (INIS)

    Nakai, Shigetsugu; Aoi, Hideki; Hayashi, Ken-ichi; Katoh, Taizo; Watanabe, Takashi.

    1988-01-01

    A gravimetric analysis using meta-boric acid (HBO 2 or DBO 2 ) as a weighing form has been developed for solubility measurement. The method gave satisfactory results in preliminary measurement of solubilities of boric acid in light water. By using this method, the solubilities of 10 B enriched D 3 BO 3 in heavy water were measured. The results are as follows; 2.67 (7deg C), 3.52 (15deg C), 5.70 (30deg C), 8.87 (50deg C) and 12.92 (70deg C) w/o, respectively. These values are about 10% lower than those in light water. Thermodynamical consideration based on the data shows that boric acid is the water structure breaker. (author)

  20. Pollution Status of Pakistan: A Retrospective Review on Heavy Metal Contamination of Water, Soil, and Vegetables

    Directory of Open Access Journals (Sweden)

    Amir Waseem

    2014-01-01

    Full Text Available Trace heavy metals, such as arsenic, cadmium, lead, chromium, nickel, and mercury, are important environmental pollutants, particularly in areas with high anthropogenic pressure. In addition to these metals, copper, manganese, iron, and zinc are also important trace micronutrients. The presence of trace heavy metals in the atmosphere, soil, and water can cause serious problems to all organisms, and the ubiquitous bioavailability of these heavy metal can result in bioaccumulation in the food chain which especially can be highly dangerous to human health. This study reviews the heavy metal contamination in several areas of Pakistan over the past few years, particularly to assess the heavy metal contamination in water (ground water, surface water, and waste water, soil, sediments, particulate matter, and vegetables. The listed contaminations affect the drinking water quality, ecological environment, and food chain. Moreover, the toxicity induced by contaminated water, soil, and vegetables poses serious threat to human health.

  1. Neutron moderation in heavy water

    International Nuclear Information System (INIS)

    Assis, J.T. de.

    1980-03-01

    The calculation of the energetic spectrum of thermic neutrons in heavy water, according to the models of the differential cross section; is presented. Simplifications in the Butler model are suggested for the diminution of computer time. The results obtained are compared with experimental data and with the Brown - St.John model. This calculation has been done in 30 energy groups and within our limit of precision, the results with the models and simplifications present satisfactory values, allowing its inclusion in reactor codes. (Author) [pt

  2. Coolant mixing in pressurized water reactors. Pt. 1. Feasibility of closed analytical solutions and simulation of the mixing with CFX-4. Final report

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Rohde, U.

    2001-10-01

    The project was aimed at the analytical and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing velocity measurements by the LDA technique. Design and construction of the ROCOM facility including the measurement equipment were performed in a second part of the project. For the design of the facility, CFD calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. A theoretical 2D-model of the downcomer flow was developed based on the potential theory. The coolant inlet is represented by mass sources. Potential vortices were superposed to describe large scale recirculations. However, the method requires an a-priory knowledge of the location and intensity of the vorticity sources. Therefore, the main goal of the project was the numerical simulation of the coolant mixing of different PWRs. The temperature and boron concentration fields established by the coolant mixing during nominal and transient flow conditions in the pressure vessel of the PWR Konvoi and the Russian type WWER-440 were investigated. The calculations were carried out with the CFD-code CFX 4. The results of the CFD calculation are found in the final report. The report is based on the Ph.D. work of T. Hoehne. (orig.) [de

  3. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  4. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  5. Heavy-Water Power Reactors. Proceedings Of A Symposium

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  6. Heavy-Water Power Reactors. Proceedings Of A Symposium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-04-15

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  7. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  8. Heavy metals in water, sediments and submerged macrophytes in ponds around the Dianchi Lake, China.

    Science.gov (United States)

    Wang, Zhixiu; Yao, Lu; Liu, Guihua; Liu, Wenzhi

    2014-09-01

    Through retaining runoff and pollutants such as heavy metals from surrounding landscapes, ponds around a lake play an important role in mitigating the impacts of human activities on lake ecosystems. In order to determine the potential for heavy metal accumulation of submerged macrophytes, we investigated the concentrations of 10 heavy metals (i.e., As, Cd, Co, Cr, Cu, Fe, Mn, Ni, Pb, and Zn) in water, sediments, and submerged macrophytes collected from 37 ponds around the Dianchi Lake in China. Our results showed that both water and sediments of these ponds were polluted by Pb. Water and sediments heavy metal concentrations in ponds received urban and agricultural runoff were not significantly higher than those in ponds received forest runoff. This result indicates that a large portion of heavy metals in these ponds may originate from atmospheric deposition and weathering of background soils. Positive relationships were found among heavy metal concentrations in submerged macrophytes, probably due to the coaccumulation of heavy metals. For most heavy metals, no significant relationships were found between submerged macrophytes and their water and sediment environments. The maximum concentrations of Cr, Fe and Ni in Ceratophyllum demersum were 4242, 16,429 and 2662mgkg(-1), respectively. The result suggests that C. demersum is a good candidate species for removing heavy metals from polluted aquatic environments. Copyright © 2014 Elsevier Inc. All rights reserved.

  9. Linear titration plot for the determination of boron in the primary coolant of a pressurized water reactor

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1992-01-01

    A linear titration plot method has been devised for the determination of boron as boric acid in partly neutralized solution, such as occurs in the primary coolant of pressurized water reactors. The total boron and the alkali in the sample are determined simultaneously. Although it is not essential to add mannitol in this method, it is more accurate when the solution is saturated with mannitol. Comparisons are made with other modes of titration: Gran plots, first and second differential potentiometric titrations and indicator titrations. None of these gives the total boron directly in partly neutralized solutions. (author)

  10. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  11. A open-quotes zero wasteclose quotes coolant management strategy

    International Nuclear Information System (INIS)

    Kennicott, M.A.

    1994-01-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant's Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant's liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study's goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a open-quotes zero wasteclose quotes machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs

  12. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  13. Lubrication analysis of the thrust bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

    2001-01-01

    Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

  14. Accelerator driven heavy water blanket on circulating fuel

    International Nuclear Information System (INIS)

    Kazaritsky, V.D.; Blagovolin, P.P.; Mladov, V.R.; Okhlopkov, M.L.; Batyaev, V.F.; Stepanov, N.V.; Seliverstov, V.V.

    1997-01-01

    A conceptual design of a heavy water blanket with circulating fuel for an accelerator driven transmutation system is described. The hybrid system consists of a high-current linear accelerator of protons and 4 targets, each placed inside a subcritical blanket

  15. Numerical Investigation on the Performance of an Automotive Thermoelectric Generator with Exhaust-Module-Coolant Direct Contact

    Science.gov (United States)

    Wang, Yiping; Tang, Yulin; Deng, Yadong; Su, Chuqi

    2018-06-01

    Energy conservation and environmental protection have typically been a concern of research. Researchers have confirmed that in automotive engines, just 12-25% of the fuel energy converts into effective work and 30-40% gets wasted in the form of exhaust. Saidur et al. (Energy Policy 37:3650, 2009) and Hasanuzzaman et al. (Energy 36:233, 2011). It will be significant to enhance fuel availability and decrease environmental pollution if the waste heat in the exhaust could be recovered. Thermoelectric generators (TEGs), which can translate heat into electricity, have become a topic of interest for vehicle exhaust waste heat recovery. In conventional automotive TEGs, the thermoelectric modules (TEMs) are arranged between the exhaust tank and the coolant tank. The TEMs do not contact the hot exhaust and coolant, which leads to low heat transfer efficiency. Moreover, to provide enough packing force to keep good contact with the exhaust tank and the coolant tank, the framework required is so robust that the TEGs become too heavy. Therefore, in current study, an automotive TEG was designed which included one exhaust channel, one coolant channel and several TEMs. In the TEG, the TEMs which contacted the exhaust and coolant directly were inserted into the walls of each coolant channel. To evaluate the performance of the automotive TEG, the flow field and temperature field were computed by computational fluid dynamics (CFD). Based on the temperature distribution obtained by CFD and the performance parameters of the modules, the total power generation was obtained by some proved empirical formulas. Compared with conventional automotive TEGs, the power generation per unit volume exhaust was boosted.

  16. Numerical Investigation on the Performance of an Automotive Thermoelectric Generator with Exhaust-Module-Coolant Direct Contact

    Science.gov (United States)

    Wang, Yiping; Tang, Yulin; Deng, Yadong; Su, Chuqi

    2017-12-01

    Energy conservation and environmental protection have typically been a concern of research. Researchers have confirmed that in automotive engines, just 12-25% of the fuel energy converts into effective work and 30-40% gets wasted in the form of exhaust. Saidur et al. (Energy Policy 37:3650, 2009) and Hasanuzzaman et al. (Energy 36:233, 2011). It will be significant to enhance fuel availability and decrease environmental pollution if the waste heat in the exhaust could be recovered. Thermoelectric generators (TEGs), which can translate heat into electricity, have become a topic of interest for vehicle exhaust waste heat recovery. In conventional automotive TEGs, the thermoelectric modules (TEMs) are arranged between the exhaust tank and the coolant tank. The TEMs do not contact the hot exhaust and coolant, which leads to low heat transfer efficiency. Moreover, to provide enough packing force to keep good contact with the exhaust tank and the coolant tank, the framework required is so robust that the TEGs become too heavy. Therefore, in current study, an automotive TEG was designed which included one exhaust channel, one coolant channel and several TEMs. In the TEG, the TEMs which contacted the exhaust and coolant directly were inserted into the walls of each coolant channel. To evaluate the performance of the automotive TEG, the flow field and temperature field were computed by computational fluid dynamics (CFD). Based on the temperature distribution obtained by CFD and the performance parameters of the modules, the total power generation was obtained by some proved empirical formulas. Compared with conventional automotive TEGs, the power generation per unit volume exhaust was boosted.

  17. Device for preventing leakage of coolant in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kobayashi, Yukio; Sekiguchi, Mamoru; Yoshida, Hideo.

    1975-01-01

    Object: To prevent leakage of coolant from between lower tie plate and channel box without causing deformation of the channel box and also without the possibility of disturbing the installation and removal of the box by the provision of a thin plate provided with leakage holes for the lower tie plate. Structure: Static water pressure within the lower tie plate is adapted to act upon the bear side of a flat plate for leakage prevention through leakage holes formed in the tie plate, thus urging the flat plate against the channel box inner surface. Meanwhile, static water pressure having been led through the leakage holes in the flat plate is adapted to press the flat plate in the vertical direction, thus urging the flat plate against the channel box inner surface and thereby preventing leakage of the coolant through a gap between the channel box and lower tie plate. (Yoshino, Y.)

  18. Halden Boiling Water Reactor. Plant Performance and Heavy-Water Management

    Energy Technology Data Exchange (ETDEWEB)

    Aas, S.; Jamne, E.; Wullum, T.; Fjellestad, K. [Institutt for Atomenergi, OECD Halden Reactor Project, Halden (Norway)

    1968-04-15

    The Halden boiling heavy-water reactor, designed and built by the Norwegian Institutt for Atomenergi, has since June 1958 been operated as an international project. On its second charge the reactor was operated at power levels up to 25 MW and most of the time at a pressure of 28.5 kg/cm{sup 2}. During the period from July 1964 to December 1966 the plant availability was close to 64% including shutdowns because of test fuel failures and loading/unloading of fuel. Disregarding such stops, the availability was close to 90%. The average burnup of the core is about 6200 MWd/t UO{sub 2} : the most highly exposed elements have reached 10000 MWd/t UO{sub 2}. The transition temperature of the reactor tank has been followed closely. The results of the surveillance programme and the implication on the reactor operation are discussed. The reactor is located in a cave in a rock. Some experiences with such a containment are given. To locate failed test-fuel elements a fuel failure location system has been installed. A fission gas collection system has saved valuable reactor time during clean-up of the reactor system following test fuel failures. Apart from one incident with two of the control stations, the plant control and instrumentation systems have functioned satisfactorily. Two incidents with losses of 150 and 200 kg of heavy water have occurred. However, after improved methods for leakage detection had been developed, the losses have been kept better than 50 g/h . Since April 1962 the isotopic purity of the heavy water (14 t) has decreased from 99.75 to 99.62%. The tritium concentration is now slightly above 700 {mu}C/cm{sup 3}. This activity level has not created any serious operational or maintenance problems. An extensive series of water chemistry experiments has been performed to study the influence of various operating parameters on radiolytic gas formation. The main results of these experiments will be reported. Different materials such as mild steel, ferritic steel

  19. Advanced technology heavy water monitors offering reduced implementation costs

    International Nuclear Information System (INIS)

    Kalechstein, W.; Hippola, K.B.

    1984-10-01

    The development of second generation heavy water monitors for use at CANDU power stations and heavy water plants has been completed and the instruments brought to the stage of commercial availability. Applications of advanced technology and reduced utilization of custom manufactured components have together resulted in instruments that are less expensive to produce than the original monitors and do not require costly station services. The design has been tested on two prototypes and fully documented, including the inspection and test procedures required for manufacture to the CSA Z299.3 quality verfication program standard. Production of the new monitors by a commercial vendor (Barringer Research Ltd.) has begun and the first instrument is scheduled for delivery to CRNL's NRU reactor in late 1984

  20. Concentration of Heavy Metals in Drinking Water from Urban Areas ...

    African Journals Online (AJOL)

    Bheema

    drinking water treatment practices in the areas, which in turn have important human health implications. This study, therefore, recommends the government and other responsible authorities to take appropriate corrective measures. Key words: Drinking water quality, Heavy metals, Maximum admissible limit, World health.

  1. Silica coated magnetite nanoparticles for removal of heavy metal ions from polluted waters

    CERN Document Server

    Dash, Monika

    2013-01-01

    Magnetic removal of Hg2+ and other heavy metal ions like Cd2+, Pb2+ etc. using silica coated magnetite particles from polluted waters is a current topic of active research to provide efficient water recycling and long term high quality water. The technique used to study the bonding characteristics of such kind of nanoparticles with the heavy metal ions is a very sensitive hyperfine specroscopy technique called the perturbed angular correlation technique (PAC).

  2. Some observations on simulated molten debris-coolant layer dynamics

    International Nuclear Information System (INIS)

    Greene, G.A.; Klein, J.; Klages, J.; Schwarz, E.; Sanborn, Y.

    1983-04-01

    Experiments are being performed to investigate high temperature liquid-liquid film boiling between a pool of liquid metal and an overlying coolant pool of R-11 or water. Film boiling has been observed to be stable for R-11; however, considerable liquid-liquid contact has been observed with water well beyond the minimum film boiling temperature. Unstable liquid-liquid film boiling of water has been observed to escalate into dispersive, non-energetic vapor explosions when the interface contact temperature exceeded the spontaneous nucleation temperature. Other parametric trends in the data are discussed

  3. Molten fuel/coolant interaction studies: some results obtained with the Windscale small shock tube rig

    International Nuclear Information System (INIS)

    Higham, E.J.; Vaughan, G.J.

    1978-02-01

    Experiments are described in which water has been brought into contact with various molten metals in a shock tube, thus simulating the fall of coolant into molten uranium dioxide in a postulated reactor accident. Impact velocities of the water on to the molten material were in the range 5 to 7 m/s. Shock-pulse pressures in the water column after impact and particle size distributions of the dispersed resolidified material that was recovered were measured. The proportion of dispersed material and the size of the shock pulse (by comparison with that expected from water hammer alone) have been used as criteria for the occurrence of a molten fuel/coolant interaction and such interactions of varying degrees of violence have been found for water/aluminium, water/bismuth, water/tin, over a range of temperatures from 350 0 C to 950 0 C, for water/boric oxide, but not for water/magnesium. (author)

  4. Heavy water production by alkaline water electrolysis

    International Nuclear Information System (INIS)

    Kamath, Sachin; Sandeep, K.C.; Bhanja, Kalyan; Mohan, Sadhana; Sugilal, G.

    2014-01-01

    Several heavy water isotope production processes are reported in literature. Water electrolysis in combination with catalytic exchange CECE process is considered as a futuristic process to increase the throughput and reduce the cryogenic distillation load but the application is limited due to the high cost of electricity. Any improvement in the efficiency of electrolyzers would make this process more attractive. The efficiency of alkaline water electrolysis is governed by various phenomena such as activation polarization, ohmic polarization and concentration polarization in the cell. A systematic study on the effect of these factors can lead to methods for improving the efficiency of the electrolyzer. A bipolar and compact type arrangement of the alkaline water electrolyzer leads to increased efficiency and reduced inventory in comparison to uni-polar tank type electrolyzers. The bipolar type arrangement is formed when a number of single cells are stacked together. Although a few experimental studies have been reported in the open literature, CFD simulation of a bipolar compact alkaline water electrolyzer with porous electrodes is not readily available.The principal aim of this study is to simulate the characteristics of a single cell compact electrolyzer unit. The simulation can be used to predict the Voltage-Current Density (V-I) characteristics, which is a measure of the efficiency of the process.The model equations were solved using COMSOL multi-physics software. The simulated V-I characteristic is compared with the experimental data

  5. Performance of refractometry in quantitative estimation of isotopic concentration of heavy water in nuclear reactor

    International Nuclear Information System (INIS)

    Dhole, K.; Roy, M.; Ghosh, S.; Datta, A.; Tripathy, M.K.; Bose, H.

    2013-01-01

    Highlights: ► Rapid analysis of heavy water samples, with precise temperature control. ► Entire composition range covered. ► Both variations in mole and wt.% of D 2 O in the heavy water sample studied. ► Standard error of calibration and prediction were estimated. - Abstract: The method of refractometry has been investigated for the quantitative estimation of isotopic concentration of heavy water (D 2 O) in a simulated water sample. Feasibility of refractometry as an excellent analytical technique for rapid and non-invasive determination of D 2 O concentration in water samples has been amply demonstrated. Temperature of the samples has been precisely controlled to eliminate the effect of temperature fluctuation on refractive index measurement. The method is found to exhibit a reasonable analytical response to its calibration performance over the purity range of 0–100% D 2 O. An accuracy of below ±1% in the measurement of isotopic purity of heavy water for the entire range could be achieved

  6. Heavy metal pollution in drinking water - a global risk for human ...

    African Journals Online (AJOL)

    Water resources in the world have been profoundly influenced over the last years by human activities, whereby the world is currently facing critical water supply and drinking water quality problems. In many parts of the world heavy metal (HM) concentrations in drinking water are higher than some international guideline ...

  7. Assessment of fiber optic sensors for aging monitoring of industrial liquid coolants

    Science.gov (United States)

    Riziotis, Christos; El Sachat, Alexandros; Markos, Christos; Velanas, Pantelis; Meristoudi, Anastasia; Papadopoulos, Aggelos

    2015-03-01

    Lately the demand for in situ and real time monitoring of industrial assets and processes has been dramatically increased. Although numerous sensing techniques have been proposed, only a small fraction can operate efficiently under harsh industrial environments. In this work the operational properties of a proposed photonic based chemical sensing scheme, capable to monitor the ageing process and the quality characteristics of coolants and lubricants in industrial heavy machinery for metal finishing processes is presented. The full spectroscopic characterization of different coolant liquids revealed that the ageing process is connected closely to the acidity/ pH value of coolants, despite the fact that the ageing process is quite complicated, affected by a number of environmental parameters such as the temperature, humidity and development of hazardous biological content as for example fungi. Efficient and low cost optical fiber sensors based on pH sensitive thin overlayers, are proposed and employed for the ageing monitoring. Active sol-gel based materials produced with various pH indicators like cresol red, bromophenol blue and chorophenol red in tetraethylorthosilicate (TEOS), were used for the production of those thin film sensitive layers deposited on polymer's and silica's large core and highly multimoded optical fibers. The optical characteristics, sensing performance and environmental robustness of those optical sensors are presented, extracting useful conclusions towards their use in industrial applications.

  8. DEGRADATION EVALUATION OF HEAVY WATER DRUMS AND TANKS

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J.; Vormelker, P.

    2009-07-31

    Heavy water with varying chemistries is currently being stored in over 6700 drums in L- and K-areas and in seven tanks in L-, K-, and C-areas. A detailed evaluation of the potential degradation of the drums and tanks, specific to their design and service conditions, has been performed to support the demonstration of their integrity throughout the desired storage period. The 55-gallon drums are of several designs with Type 304 stainless steel as the material of construction. The tanks have capacities ranging from 8000 to 45600 gallons and are made of Type 304 stainless steel. The drums and tanks were designed and fabricated to national regulations, codes and standards per procurement specifications for the Savannah River Site. The drums have had approximately 25 leakage failures over their 50+ years of use with the last drum failure occurring in 2003. The tanks have experienced no leaks to date. The failures in the drums have occurred principally near the bottom weld, which attaches the bottom to the drum sidewall. Failures have occurred by pitting, crevice and stress corrosion cracking and are attributable, in part, to the presence of chloride ions in the heavy water. Probable degradation mechanisms for the continued storage of heavy water were evaluated that could lead to future failures in the drum or tanks. This evaluation will be used to support establishment of an inspection plan which will include susceptible locations, methods, and frequencies for the drums and tanks to avoid future leakage failures.

  9. AECB staff annual report of Bruce Heavy Water Plant operation for the year 1991

    International Nuclear Information System (INIS)

    1992-11-01

    Bruce Heavy Water Plant operation was acceptably safe in 1991. There were no breaches of any of the regulations issued under the authority of the Atomic Energy Control Act. There was one violation of the operating licence. For one hour on October 30, 1991, water leaving the plant contained more hydrogen sulphide than Ontario regulations allow. There was no threat to public health or safety or harm to the environment as a result of this violation. One worker was overcome by hydrogen sulphide. The worker did not lose consciousness, but had the symptoms of H 2 S poisoning. Ontario Hydro took actions to increase awareness of the Operating Policy and Principles at Bruce Heavy Water Plant during 1991. All personnel attended a training course, and Ontario Hydro is reviewing all Bruce Heavy Water Plant documentation to ensure it is consistent with the Operating Policies and Principles. Ontario Hydro met 13 of 15 safety-related system availability targets. The AECB is satisfied appropriate action is being taken to improve the performance of the other two systems. Ontario Hydro continued to put heavy emphasis on safety training; however, they did not meet some of their other training targets. Ontario Hydro completed all of the planned emergency exercises at Bruce Heavy Water Plant in 1991. (Author)

  10. Nanofluid as coolant for grinding process: An overview

    Science.gov (United States)

    Kananathan, J.; Samykano, M.; Sudhakar, K.; Subramaniam, S. R.; Selavamani, S. K.; Manoj Kumar, Nallapaneni; Keng, Ngui Wai; Kadirgama, K.; Hamzah, W. A. W.; Harun, W. S. W.

    2018-04-01

    This paper reviews the recent progress and applications of nanoparticles in lubricants as a coolant (cutting fluid) for grinding process. The role of grinding machining in manufacturing and the importance of lubrication fluids during material removal are discussed. In grinding process, coolants are used to improve the surface finish, wheel wear, flush the chips and to reduce the work-piece thermal deformation. The conventional cooling technique, i.e., flood cooling delivers a large amount of fluid and mist which hazardous to the environment and humans. Industries are actively looking for possible ways to reduce the volume of coolants used in metal removing operations due to the economical and ecological impacts. Thus as an alternative, an advanced cooling technique known as Minimum Quantity Lubrication (MQL) has been introduced to the enhance the surface finish, minimize the cost, to reduce the environmental impacts and to reduce the metal cutting fluid consumptions. Nanofluid is a new-fangled class of fluids engineered by dispersing nanometre-size solid particles into base fluids such as water, lubrication oils to further improve the properties of the lubricant or coolant. In addition to advanced cooling technique review, this paper also reviews the application of various nanoparticles and their performance in grinding operations. The performance of nanoparticles related to the cutting forces, surface finish, tool wear, and temperature at the cutting zone are briefly reviewed. The study reveals that the excellent properties of the nanofluid can be beneficial in cooling and lubricating application in the manufacturing process.

  11. Significant Features of Warm Season Water Vapor Flux Related to Heavy Rainfall and Draught in Japan

    Science.gov (United States)

    Nishiyama, Koji; Iseri, Yoshihiko; Jinno, Kenji

    2009-11-01

    In this study, our objective is to reveal complicated relationships between spatial water vapor inflow patterns and heavy rainfall activities in Kyushu located in the western part of Japan, using the outcomes of pattern recognition of water vapor inflow, based on the Self-Organizing Map. Consequently, it could be confirmed that water vapor inflow patterns control the distribution and the frequency of heavy rainfall depending on the direction of their fluxes and the intensity of Precipitable water. Historically serious flood disasters in South Kyushu in 1993 were characterized by high frequency of the water vapor inflow patterns linking to heavy rainfall. On the other hand, severe draught in 1994 was characterized by inactive frontal activity that do not related to heavy rainfall.

  12. Impact analysis and testing of tritiated heavy water transportation packages including hydrodynamic effects

    International Nuclear Information System (INIS)

    Sauve, R.G.; Tulk, J.D.; Gavin, M.E.

    1989-01-01

    Ontario Hydro has recently designed a new Type B(M) Tritiated Heavy Water Transportation Package (THWTP) for the road transportation of tritiated heavy water from its operating nuclear stations to the Tritium Removal Facility in Ontario. These packages must demonstrate the ability to withstand severe shock and impact scenarios such as those prescribed by IAEA standards. The package, shown in figure 1, comprises an inner container filled with tritiated heavy water, and a 19 lb/ft 3 polyurethane foam-filled overpack. The overpack is of sandwich construction with 304L stainless steel liners and 10.5 inch thick nominal foam walls. The outer shell is 0.75 inch thick and the inner shell is 0.25 inch thick. The primary containment boundary consists of the overpack inner liner, the containment lid and outer containment seals in the lid region. The total weight of the container including the 12,000 lb. payload is 36,700 lb. The objective of the present study is to evaluate the hydrodynamic effect of the tritiated heavy water payload on the structural integrity of the THWTP during a flat end drop from a height of 9 m. The study consisted of three phases: (i) developing an analytical model to simulate the hydrodynamic effects of the heavy water payload during impact; (ii) performing an impact analysis for a 9 m flat end drop of the THWTP including fluid structure interaction; (iii) verification of the analytical models by experiment

  13. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    International Nuclear Information System (INIS)

    DeVan, J.H.

    1979-01-01

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF-BeF 2 , Pb-Li alloys, and solid ceramic compounds such as Li 2 O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies. (orig.)

  14. Fission Product Releases from a Core into a Coolant of a Prismatic 350-MWth HTR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Min; Jo, C. K. [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    A prismatic 350-MW{sub th} high temperature reactor (HTR) is a means to generate electricity and process heat for hydrogen production. The HTR will be operated for an extended fuel burnup of more than 150 GWd/MTU. Korea Atomic Energy Research Institute (KAERI) is performing a point design for the HTR which is a pre-conceptual design for the analysis and assessment of engineering feasibility of the reactor. In a prismatic HTR, metallic and gaseous fission products (FPs) are produced in the fuel, moved through fuel materials, and released into a primary coolant. The FPs released into the coolant are deposited on the various helium-wetted surfaces in the primary circuit, or they are sorbed on particulate matters in the primary coolant. The deposited or sorbed FPs are released into the environment through the leakage or venting of the primary coolant. It is necessary to rigorously estimate such radioactivity releases into the environment for securing the health and safety of the occupational personnel and the public. This study treats the FP releases from a core into a coolant of a prismatic 350-MW{sub th} HTR. These results can be utilized as input data for the estimation of FP migration from a coolant into the environment. The analysis of fission product release within a prismatic 350-MW{sub th} HTR has been done. It was assumed that the HTR was operated at constant temperature and power for 1500 EFPDs. - The final burnup is 152 GWd/tHM at packing fraction of 25 %, and the final fast fluence is about 8 X 10{sup 21} n/cm{sup 2}, E{sub n} > 0.1 MeV. - The temperatures at the compact center and at the center of a kernel located at the compact center are 884 and 893 .deg. C, respectively, when the packing fraction is 25 % and the coolant temperature is 850 .deg. C. - Xenon is the most radioactive fission product in a coolant of a prismatic HTR when there are broken TRISOs and fuel component contaminated with heavy metals. For metallic fission products, the radioactivity

  15. Determination of primary flow by correlation of temperatures of the coolant

    International Nuclear Information System (INIS)

    Villanueva, Jose

    2003-01-01

    Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

  16. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  17. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  18. Deposition and incorporation of corrosion product to primary coolant suppressing method

    International Nuclear Information System (INIS)

    Tsuzuki, Yasuo; Hasegawa, Naoyoshi; Fujioka, Tsunaaki.

    1992-01-01

    In a PWR type nuclear power plant, the concentration of dissolved nitrogen in primary coolants is increased by controlling the nitrogen partial pressure in a volume controlling tank gas phase portion or addition of water in a primary system water supply tank containing dissolved nitrogen to a primary system. Then ammonium is formed by a reaction with hydrogen dissolved in the primary coolants in the field of radiation rays, to control the concentration of ammonium in the coolants within a range from 0.5 to 3.5 ppm, and operate the power plant. As a result, deposition and incorporation of corrosion products to the structural materials of the primary system equipments during plant operation (pH 6.8 to 8.0) are suppressed. In other words, deposition of particulate corrosion products on the surface of fuel cladding tubes and the inner surface of pipelines in the primary system main equipments is prevented and incorporation of ionic radioactive corrosion products to the oxide membranes on the inner surface of the pipelines of the primary system main equipments is suppressed, to greatly reduce the radiation dose rate of the primary system pipelines. Thus, operator's radiation exposure can be decreased upon shut down of the plant. (N.H.)

  19. Tritium separation from heavy water by electrolysis with solid polymer electrolyte

    International Nuclear Information System (INIS)

    Ogata, Y.; Ohtani, N.; Kotaka, M.

    2003-01-01

    A tritium separation from heavy water by electrolysis using a solid polymer electrode layer was specified. The cathode was made of stainless steel or nickel. The electrolysis was performed for 1 hour at 5, 10, 20, and 30 deg C. Using a palladium catalyst, generated hydrogen and oxygen gases were recombined, which was collected with a cold trap. The activities of the samples were measured by a liquid scintillation counter. The apparent tritium separation factors of the heavy and light water at 20 deg C were ∼2 and ∼12, respectively. (author)

  20. Improvements done at Heavy Water Plant (Manuguru) to increase the standards of environmental protection

    International Nuclear Information System (INIS)

    Rama Rao, V.V.S.; Gupta, R.V.; Pandey, B.L.

    1997-01-01

    The Heavy Water Plant at Manuguru is designed to produce 185 MTY of nuclear grade heavy water based on bithermal H 2 S-H 2 O exchange process and handles large inventory of H 2 S gas (about 400 MT). As H 2 S gas is very toxic, corrosive and hazardous in nature, extreme care has been taken in the design of plant, selection of equipment and materials adhering to stringent fabrication procedures and codes to ensure the production of heavy water in a safe manner. This paper highlights the improvements done at Heavy Water Plant (Manuguru) to increase the standards of environmental protection. The safety assessment of a hazardous plant is a continuous process. Apart from the extreme care taken in the design, construction, commissioning and operation of the plant, review of each and every safety related unusual occurrence by various levels of review committees as stipulated and speedy implementation of the recommendations goes in a long way in increasing the standards of environmental protection