Secondary coolant purification system

International Nuclear Information System (INIS)

Stiteler, F.Z.; Donohue, J.P.

1978-01-01

The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

International Nuclear Information System (INIS)

Liebig, E.

1977-01-01

A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

Secondary coolant circuit operation tests: steam generator feedwater supply

International Nuclear Information System (INIS)

Beroux, M.

1985-01-01

No one important accident occurred during the start-up tests of the 1300MWe P4 series, concerning the feedwater system of steam generators (SG). This communication comments on some incidents, that the tests allowed to detect very soon and which had no consequences on the operation of units: 1) Water hammer in feedwater tubes, and incidents met in the emergency steam generator water supply circuit. The technological differences between SG 900 and 1300 are pointed out, and the measures taken to prevent this problem are presented. 2) Incidents met on the emergency feedwater supply circuit of steam generators; mechanical or functional modifications involved by these incidents [fr

Hard alloys testing-machine for values of PWR primary coolant circuits

International Nuclear Information System (INIS)

Campan, J.L.; Sauze, A.

1980-01-01

Testing of valve parts or material used in valve fabrication and particularly seizing conditions in friction of plane surfaces coated with hard alloys of the type stellite. The testing equipment called Marguerite is composed of a hot pressurized water loop in conditions similar to PWR primary coolant circuits (320 0 C, 150 bars) and a testing-machine with measuring instruments. Testing conditions and samples are described [fr

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)

The impact of radiolytic yield on the calculated ECP in PWR primary coolant circuits

International Nuclear Information System (INIS)

Urquidi-Macdonald, Mirna; Pitt, Jonathan; Macdonald, Digby D.

2007-01-01

A code, PWR-ECP, comprising chemistry, radiolysis, and mixed potential models has been developed to calculate radiolytic species concentrations and the corrosion potential of structural components at closely spaced points around the primary coolant circuits of pressurized water reactors (PWRs). The pH(T) of the coolant is calculated at each point of the primary-loop using a chemistry model for the B(OH) 3 + LiOH system. Although the chemistry/radiolysis/mixed potential code has the ability to calculate the transient reactor response, only the reactor steady state condition (normal operation) is discussed in this paper. The radiolysis model is a modified version of the code previously developed by Macdonald and coworkers to model the radiochemistry and corrosion properties of boiling water reactor primary coolant circuits. In the present work, the PWR-ECP code is used to explore the sensitivity of the calculated electrochemical corrosion potential (ECP) to the set of radiolytic yield data adopted; in this case, one set had been developed from ambient temperature experiments and another set reported elevated temperatures data. The calculations show that the calculated ECP is sensitive to the adopted values for the radiolytic yields

Monitoring of coolant temperature stratification on piping components in WWER-440 NPPs

International Nuclear Information System (INIS)

Hudcovsky, S.; Slanina, M.; Badiar, S.

2001-01-01

The presentation deals with the aims of non-standard temperature measurements installed on primary and secondary circuit in WWER-440 NPPs, explains reasons of coolant temperature stratification on the piping components. It describes methods of the measurements on pipings, range of installation of the temperature measurements in EBO and EMO units and illustrates results of measurements of coolant temperature stratification. (Authors)

Approach to normalization of the secondary circuit water chemistry of NPP with WWER-1000

International Nuclear Information System (INIS)

Mamet, V.A.; Erpyleva, S.F.; Banyuk, G.F.

1998-01-01

The approach to normalization if indices of water-chemical regime of the secondary circuit of the NPP with WWER-1000 reactor, based on pH calculational values at the coolant working temperature in dependence on the normalized admixtures concentration is considered. The possibility for conducting the water regime of steam generators by the ratio of sodium concentration and electrical conductivity of H-cation sample of blow-through water is shown. The limitations (os action level) by deviation of normalized indices from recommended ones for normal operational conditions are described

Modeling of corrosion product migration in the secondary circuit of nuclear power plants with WWER-1200

Science.gov (United States)

Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.; Motkova, E. A.; Zelenina, E. V.; Prokhorov, N. A.; Gorbatenko, S. P.; Tsitser, A. A.

2016-04-01

Models of corrosion and mass transfer of corrosion products in the pipes of the condensate-feeding and steam paths of the secondary circuit of NPPs with WWER-1200 are presented. The mass transfer and distribution of corrosion products over the currents of the working medium of the secondary circuit were calculated using the physicochemical model of mass transfer of corrosion products in which the secondary circuit is regarded as a cyclic system consisting of a number of interrelated elements. The circuit was divided into calculated regions in which the change in the parameters (flow rate, temperature, and pressure) was traced and the rates of corrosion and corrosion products entrainment, high-temperature pH, and iron concentration were calculated. The models were verified according to the results of chemical analyses at Kalinin NPP and iron corrosion product concentrations in the feed water at different NPPs depending on pH at 25°C (pH25) for service times τ ≥ 5000 h. The calculated pH values at a coolant temperature t (pH t ) in the secondary circuit of NPPs with WWER-1200 were presented. The calculation of the distribution of pH t and ethanolamine and ammonia concentrations over the condensate feed (CFC) and steam circuits is given. The models are designed for developing the calculation codes. The project solutions of ATOMPROEKT satisfy the safety and reliability requirements for power plants with WWER-1200. The calculated corrosion and corrosion product mass transfer parameters showed that the model allows the designer to choose between the increase of the correcting reagent concentration, the use of steel with higher chromium contents, and intermittent washing of the steam generator from sediments as the best solution for definite regions of the circuit.

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)

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.)

Experience in vibro-acoustic control of primary coolant circuit aggregates

International Nuclear Information System (INIS)

Sedov, V.K.; Adamenkov, K.A.

1977-01-01

Fundamental principles and possibilities of vibro-acoustic control of the primary coolant circuit in nuclear power plants for detecting failures (slack parts, penetration of foreign bodies, crack formation, etc.) are presented. As a result of pressure and flow rate fluctuations such failures give rise to characteristic changes in apmplitude and frequency of vibration and technological noise from the different aggregates with respect to a 'calibration' spectrum taken in the intact state. Nature and location of the failures may be determined by statistical analysis of the signals recorded from pressure and acceleration gauges. Certain parts of the primary circuit are controlled, especially the main circulation pumps. Additionally, neutron noise has been measured in order to control the core insertions. The method is illustrated by means of measurements performed in the units 1 to 4 of the Novovoronezh nuclear power plant during start-up operation and continuous operation. (author)

Experience in vibro-acoustic control of primary coolant circuit aggregates

Energy Technology Data Exchange (ETDEWEB)

Sedov, V K; Adamenkov, K A [Nuclear power plant Novo-Voronesh (USSR)

1977-10-01

Fundamental principles and possibilities of vibro-acoustic control of the primary coolant circuit in nuclear power plants for detecting failures (slack parts, penetration of foreign bodies, crack formation, etc.) are presented. As a result of pressure and flow rate fluctuations such failures give rise to characteristic changes in apmplitude and frequency of vibration and technological noise from the different aggregates with respect to a 'calibration' spectrum taken in the intact state. Nature and location of the failures may be determined by statistical analysis of the signals recorded from pressure and acceleration gauges. Certain parts of the primary circuit are controlled, especially the main circulation pumps. Additionally, neutron noise has been measured in order to control the core insertions. The method is illustrated by means of measurements performed in the units 1 to 4 of the Novovoronezh nuclear power plant during start-up operation and continuous operation.

Conceptual design of the integral test loop (I): Reactor coolant system and secondary system

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Lee, Seong Je; Kwon, Tae Soon; Moon, Sang Ki [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-10-01

This report describes the conceptual design of the primary coolant system and the secondary system of the Integral Test Loop (ITL) which simulates overall thermal hydraulic phenomena of the primary system of a nuclear power plant during postulated accidents or transients. The design basis for the primary coolant system and secondary system is as follows ; Reference plant: Korean Standard Nuclear Plant (KSNP), Height ratio : 1/1, Volume ratio : 1/200, Power scale : Max. 15% of the scaled nominal power, Temperature, Pressure : Real plant conditions. The primary coolant system includes a reactor vessel, which contains a core simulator, a steam generator, a reactor coolant pump simulator, a pressurizer and piping, which consists of two hot legs, four cold legs and four intermediate legs. The secondary system consists of s steam discharge system, a feedwater supply system and a steam condensing system. This conceptual design report describes general configuration of the reference plant, and major function and operation of each system of the plant. Also described is the design philosophy of each component and system of the ITL, and specified are the design criteria and technical specifications of each component and system of the ITL in the report. 17 refs., 43 figs., 51 tabs. (Author)

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)

Integrated equipment for increasing and maintaining coolant pressure in primary circuit of PWR nuclear power plant

International Nuclear Information System (INIS)

Sykora, D.

1986-01-01

An open heat pump circuit is claimed connected to the primary circuit. The pump circuit consists of a steam pressurizer with a built-in steam distributor, a compressor, an expander, a reducing valve, an auxiliary pump, and of water and steam pipes. The operation is described and a block diagram is shown of integrated equipment for increasing and maintaining pressure in the nuclear power plant primary circuit. The appropriate entropy diagram is also shown. The advantage of the open pump circuit consists in reducing the electric power input and electric power consumption for the steam pressurizers, removing entropy loss in heat transfer with high temperature gradient, in the possibility of inserting, between the expander and the auxiliary pump, a primary circuit coolant treatment station, in simplified design and manufacture of the high-pressure steam pressurizer vessel, reducing the weight of the steam pressurizer by changing its shape from cylindrical to spherical, increasing the rate of pressure growth in the primary circuit. (E.S.)

Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

International Nuclear Information System (INIS)

Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V.; Korolev, Alexander S.

2017-01-01

The JSC ''Atomtechenergo'' experts have developed a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants. The necessity for developing the new methodology was determined by the need to decrease the calculation error of the weighted mean coolant temperature in the hot legs because of the coolant temperature stratification. The methodology development was based on the findings of experimental and calculating research executed by the authors. The methodology verification was fulfilled through comparison of calculation results obtained with and without the methodology use in various operational states and modes of several WWER-1000 power units. The obtained verification results have confirmed that the use of the new methodology provides objective error decrease in determining the weighted mean coolant temperature in the primary circuit hot legs. The decrease value depends on the stratification character which is various for different objects and conditions.

Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants

Energy Technology Data Exchange (ETDEWEB)

Saunin, Yuri V.; Dobrotvorski, Alexander N.; Semenikhin, Alexander V.; Korolev, Alexander S. [JSC ' ' Atomtechenergo' ' , Novovoronezh (Russian Federation). Novovoronezh Filial ' ' Novovoronezhatomtechenergo' ' ; Ryasny, Sergei I. [JSC ' ' Atomtechenergo' ' , Moscow (Russian Federation)

2017-09-15

The JSC ''Atomtechenergo'' experts have developed a new methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants. The necessity for developing the new methodology was determined by the need to decrease the calculation error of the weighted mean coolant temperature in the hot legs because of the coolant temperature stratification. The methodology development was based on the findings of experimental and calculating research executed by the authors. The methodology verification was fulfilled through comparison of calculation results obtained with and without the methodology use in various operational states and modes of several WWER-1000 power units. The obtained verification results have confirmed that the use of the new methodology provides objective error decrease in determining the weighted mean coolant temperature in the primary circuit hot legs. The decrease value depends on the stratification character which is various for different objects and conditions.

Secondary seal effects in hydrostatic non-contact seals for reactor coolant pump shaft

International Nuclear Information System (INIS)

Fujita, T.; Koga, T.; Tanoue, H.; Hirabayashi, H.

1987-01-01

The paper presents a seal flow analysis in a hydrostatic non-contact seal for a PWR coolant pump shaft. A description is given of the non-contact seal for the reactor coolant pump. Results are presented for a distortion analysis of the seal ring, along with the seal flow characteristics and the contact pressure profiles of the secondary seals. The results of the work confirm previously reported findings that the seal ring distortion is sensitive to the o-ring location (which was placed between the ceramic seal face and the seal ring retainer). The paper concludes that the seal flow characteristics and the tracking performance depend upon the dynamic properties of the secondary seal. (U.K.)

Studies of iodine adsorption and desorption on HTGR coolant circuit materials

International Nuclear Information System (INIS)

Osborne, M.F.; Compere, E.L.; de Nordwall, H.J.

1976-04-01

Safety studies of the HTGR system indicate that radioactive iodine, released from the fuel to the helium coolant, may pose a problem of concern if no attenuation of the amount of iodine released occurs in the coolant circuit. Since information on iodine behavior in this system was incomplete, iodine adsorption on HTGR materials was studied in vacuum as a function of iodine pressure and of adsorber temperature. Iodine coverages on Fe 3 O 4 and Cr 2 O 3 approached maxima of about 2 x 10 14 and 1 x 10 14 atoms/cm 2 , respectively, whereas the iodine coverage on graphite under similar conditions was found to be less by a factor of about 100. Iodine desorption from the same materials into vacuum or flowing helium was investigated, on a limited basis, as a function of iodine coverage, of adsorber temperature, and of dry vs wet helium. The rate of vacuum desorption from Fe 3 O 4 was related to the spectrum of energies of the adsorption sites. A small amount of water vapor in the helium enhanced desorption from iron powder but appeared to have less effect on desorption from the metal oxides

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.)

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)

Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

2011-01-01

Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

33 CFR 183.440 - Secondary circuits of ignition systems.

Science.gov (United States)

2010-07-01

... systems. 183.440 Section 183.440 Navigation and Navigable Waters COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Electrical Systems Manufacturer Requirements § 183.440 Secondary circuits of ignition systems. (a) Each conductor in a secondary circuit of an...

Secondary Circuit Start Up Chemistry Optimisation

International Nuclear Information System (INIS)

Fontan, Guillaume; Morel, Pascal

2012-09-01

In a context of investment and renewal of equipment, Electricite De France (EDF) put enhanced efforts on operating practices during start-up of the secondary circuit, in order to improve operational performance and materials lifetime. This article focuses on the objective of optimizing the filling, the chemical conditioning and the thermal conditioning of the secondary fluid, while taking into account the following issues: - Limiting the time required to obtain a proper chemistry, - Limiting the amount of water and steam used, - Limiting the amount of effluent generated. The scope is all start-up conditions of secondary circuit, both after refuelling outage or fortuitous shutdowns of the plant. The recommendations produced are based on existing local procedures and good practices, which were collected and developed in order to propose a generic methodology understandable and useful both for operators, chemists and managers. (authors)

The electrochemistry of IGSCC mitigation in BWR coolant circuits

International Nuclear Information System (INIS)

Macdonald, D.D.

2002-01-01

A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O 2 , H 2 , H 2 O 2 ), external solution composition (concentrations of O 2 , H 2 O 2 , and H 2 ), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

The electrochemistry of IGSCC mitigation in BWR coolant circuits

Energy Technology Data Exchange (ETDEWEB)

Macdonald, D.D. [Center for Electrochemical Science and Technology, The Pennsylvania State Univ., University Park, PA (United States)

2002-07-01

A brief review is presented of the electrochemical mitigation of IGSCC in water-cooled reactor heat transport circuit structural materials. Electrochemical control and mitigation is possible, because of the existence of a critical potential for IGSCC and by the feasibility of modifying the environment to displace the corrosion potential (ECP) to a value that is more negative than the critical value. However, even in cases where the ECP cannot be displaced sufficiently in the negative direction to become more negative than the critical potential, considerable advantage is accrued, because of the roughly exponential dependence of crack growth rate on potential. The most important parameters in affecting electrochemical control over the ECP and crack growth rate are the kinetic parameters (exchange current densities and Tafel constants) for the redox reactions involving the principal radiolysis products of water (O{sub 2}, H{sub 2}, H{sub 2}O{sub 2}), external solution composition (concentrations of O{sub 2}, H{sub 2}O{sub 2}, and H{sub 2}), flow velocity, and the conductivity of the bulk environment. The kinetic parameters for the redox reactions essentially determine the charge transfer impedance of the steel surface, which is shown to be one of the key parameters in affecting the magnitude of the coupling current and hence the crack growth rate. The exchange current densities, in particular, are amenable to control by catalysis or inhibition, with the result that surface modification techniques are highly effective in controlling and mitigating IGSCC in reactor coolant circuit materials. (authors)

Filtering device for primary coolant circuits in BWR type reactors

International Nuclear Information System (INIS)

Tajima, Fumio; Yamamoto, Tetsuo.

1985-01-01

Purpose: To obtain a filtering device with a large filtering area and requiring less space. Constitution: A condensate inlet for introducing condensates to be filtered of primary coolant circuits, a filtrate exit, a backwash water exit and a bent tube are disposed to a container, and a plurality of hollow thread membrane modules are suspended in the container. The condensates are caused to flow through the condensate inlet, filtered through the hollow thread membrane and then discharged from the filtrate exit. When the filtering treatment is proceeded to some extent, since solid contents captured in the hollow thread membranes are accumulated, a differential pressure is produced between the condensate inlet and the filtrate exit. When the differential pressure reaches a predetermined value, the backwash is conducted to discharge the liquid cleaning wastes through the backwash exit. The bent tube disposed to the container body is used for water and air draining. The hollow thread membranes are formed with porous resin such as of polyethylene. (Kawakami, Y.)

BR-5 primary circuit decontamination

International Nuclear Information System (INIS)

Efimov, I.A.; Nikulin, M.P.; Smirnov-Averin, A.P.; Tymosh, B.S.; Shereshkov, V.S.

1976-01-01

Results and methodology of steam-water and acid decontamination of the primary coolant circuit SBR-5 reactor in 1971 are discussed. Regeneration process in a cold trap of the primary coolant circuit is discussed

Radiation leakage monitoring method and device from primary to secondary coolant systems in nuclear reactor

International Nuclear Information System (INIS)

Tajiri, Yoshiaki; Umehara, Toshihiro; Yamada, Masataka.

1993-01-01

The present invention monitors radiation leaked from any one of primary cooling systems to secondary cooling systems in a plurality of steam generators. That is, radiation monitoring means each corresponding to steam each generators are disposed to the upstream of a position where main steam pipes are joined. With such a constitution, since the detection object of each of radiation monitoring means is secondary coolants before mixing with secondary coolants of other secondary loops or dilution, lowering of detection accuracy can be avoided. Except for the abnormal case, that is, a case neither of radiation leakage nor of background change, the device is adapted as a convenient measuring system only with calculation performance. Once abnormality occurs, a loop having a value exceeding a standard value is identified by a single channel analyzer function. The amount of radiation leakage from the steam generator belonging to the specified loop is monitored quantitatively by a multichannel analyzer function. According to the method of the present invention, since specific spectrum analysis is conducted upon occurrence of abnormality, presence of radiation leakage and the scale thereof can be judged rapidly. (I.S.)

New cooling system of the FRG-1 two barrier system of the primary coolant cycle

International Nuclear Information System (INIS)

Knop, W.; Schreiner, P.

2003-01-01

The GKSS research center operates the swimming pool reactor FRG-1 with a thermal power of 5 MW as national neutron source for neutron scattering experiments and sample irradiation as well. Before changing the primary coolant cycle consisted of the reactor core and the closed piping including pumps, heat exchanger and delay tank. The closed cooling circuit was located underneath the reactor pool, in the so-called radioactive cellar. This piping system served secondary coolant system. Due to the location of the primary coolant cycle below the operation pool a postulated 2-F line break and simultaneous failure of the pool slide gate valve could lead to a falling dry of the total reactor core. the new primary coolant system was built in the beginning 2002 in a partitioned cell all within the radioactive cellar, so that the reactor core remains with water with the assumed incident. Due to the new two barrier-inclusion of the primary circuit only the melting of two fuel plates (from total 252 fuel plates) has to be taken into account. This measure and the core compactness in 2000 with a neutron flux gain of a factor of 2 makes the FRG-1 ready for the next 15 years of reactor operation. (author)

Secondary School Students' Misconceptions about Simple Electric Circuits

Science.gov (United States)

Küçüközer, Hüseyin; Kocakülah, Sabri

2007-01-01

The aim of this study is to reveal secondary school students' misconceptions about simple electric circuits and to define whether specific misconceptions peculiar to Turkish students exist within those identified. Data were obtained with a conceptual understanding test for simple electric circuits and semi-structured interviews. Conceptual…

Increasing of prediction reliability of calcium carbonate scale formation in heat exchanger of secondary coolant circuits of thermal and nuclear power plants

International Nuclear Information System (INIS)

Tret'yakov, O.V.; Kritskij, V.G.; Styazhkin, P.S.

1991-01-01

Calcium carbonate scale formation in the secondary circuit heat exchanger of thermal and nuclear power plants is investigated. A model of calcium-carbonate scale formation providing quite reliable prediction of process running and the possibility of its control affecting the parameters of hydrochemical regime (HCR) is developed. The results can be used when designing the automatic-control system of HCR

Thermal-hydraulic model of the primary coolant circuits for the full-scale training facility with WWER-1000

International Nuclear Information System (INIS)

Kroshilin, A.E.; Zhukavin, A.P.; Pryakhin, V.N.

1992-01-01

The mathematical model realized in the full-scale educational facility for NPP operator training is described. The RETACT computational complex providing real time process simulation for all regimes including the maximum credible accident is used for calculation of thermohydraulic parameters of the primary coolant circuits and steam generator under stationary and transient conditions. The two-velocity two-temperature model of one-dimensional steam-water flow containing uncondensed gases is realized in the program

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

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

Effect of high-temperature filtration on impurity composition in the primary circuit coolant of power units with WWER-1000 reactors

International Nuclear Information System (INIS)

Efimov, A.A.; Moskvin, L.N.; Gusev, B.A.; Leont'ev, G.G.; Nekrest'yanov, S.N.

1992-01-01

The effects of high-temperature filtration on changes in dispersive, chemical, radioisotope and phase compositions of impurities in primary circuit coolant of NPP with the WWER-1000 reactor are studied. Special filters are used for the studies. The data obtained confirm the applicability of high-temperature filtration for purification of WWER reactor water and steam separators at NPPs with RBMK reactors

Method for determining the optimum mode of operation of the chemical water regime in the water-steam-circuit of power plants

International Nuclear Information System (INIS)

Sommerfeldt, P.; Reisner, H.; Hartmann, G.; Kulicke, P.

1988-01-01

The method aims at increasing the lifetime of secondary coolant circuit components in nuclear power plants through the determination of the optimum mode of operation of the chemical water regime by help of radioisotopes

Engine Tune-Up Service. Unit 4: Secondary Circuit. Posttests. Automotive Mechanics Curriculum.

Science.gov (United States)

Morse, David T.

This book of posttests is designed to accompany the Engine Tune-Up Service Student Guide for Unit 4, Secondary Circuit, available separately as CE 031 214. Focus of the posttests is testing and servicing the secondary ignition circuit. One multiple choice posttest is provided that covers the seven performance objectives contained in the unit. (No…

Nuclear Power Plants Secondary Circuit Piping Wall-Thinning Management in China

International Nuclear Information System (INIS)

Zhong Zhimin; Li Jinsong; Zheng Hui

2012-01-01

Research and field feedbacks showed that nuclear power plants secondary circuit steam and water piping are more sensitive than that of fuel plant to the attack of flow-accelerated corrosion (FAC). FAC, Liquid droplet impingement or cavitation erosion will cause secondary circuit piping local wall-thinning in NPPs. Without effective management, the wall-thinning in those high energy piping will cause leakage or pipe rupture during nuclear power plant operation, more seriously cause unplanned shut down, injured and fatality, or heavy economic losses. This paper briefly introduces the history, development and state of the art of secondary circuit piping wall-thinning management in China NPPs. Then, the effectiveness of inspection grid size selecting was analyzed in detail based on field feedbacks. EPRI recommendatory inspection grid, JSME code recommendatory grid and plant specific inspection grid were compared and the detection probabilities of local wall-thinning were estimated. Then, the development and application of NPPs Secondary Circuit Piping Wall Thickness Management Information System, developed, operated and maintained by our team, was briefly introduced and the statistical analysis results of 11 PWR units were shared. It was conclude that the long term, systemic, effective wall-thinning management strategy of high energy piping was very important to the safety and economic operation of NPPs. Furthermore, take into account the actual situation of China nuclear power plants, some advice and suggestion on developing effective nuclear power plant secondary circuit steam and water piping wall-thinning management system are put forward from code development, design and manufacture, operation management, pipeline and locations selection, inspection method selection and application, thickness measurement result evaluation, residual life predication and decision making, feedbacks usage, personnel training and etc. (author)

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.)

Coolant rate distribution in horizontal steam generator under natural circulation

International Nuclear Information System (INIS)

Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

1997-01-01

In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered

Coolant rate distribution in horizontal steam generator under natural circulation

Energy Technology Data Exchange (ETDEWEB)

Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

1997-12-31

In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

Coolant rate distribution in horizontal steam generator under natural circulation

Energy Technology Data Exchange (ETDEWEB)

Blagovechtchenski, A; Leontieva, V; Mitrioukhin, A [St. Petersburg State Technical Univ. (Russian Federation)

1998-12-31

In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

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

IEA-R1 primary and secondary coolant piping systems coupled stress analysis

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A.; Mattar Neto, Miguel

2013-01-01

The aim of this work is to perform the stress analysis of a coupled primary and secondary piping system of the IEA-R1 based on tridimensional model, taking into account the as built conditions. The nuclear research reactor IEA-R1 is a pool type reactor projected by Babcox-Willcox, which is operated by IPEN since 1957. The operation to 5 MW power limit was only possible after the conduction of life management and modernization programs in the last two decades. In these programs the components of the coolant systems, which are responsible for the water circulation into the reactor core to remove the heat generated inside it, were almost totally refurbished. The changes in the primary and secondary systems, mainly the replacement of pump and heat-exchanger, implied in piping layout modifications, and, therefore, the stress condition of the piping systems had to be reanalyzed. In this paper the structural stress assessment of the coupled primary and secondary piping systems is presented and the final results are discussed. (author)

Qualitative infrared spectral analysis of products adsorbed by silica gel from ditolylmethane coolant and their adsorption isotherm

International Nuclear Information System (INIS)

Ermakov, V.A.; Benderskaya, O.S.

1987-01-01

The IR-spectral analysis has been applied to study the products adsorbed from ditolylmethane first-circuit coolant, as well as from still bottoms after coolant distillation on silicagel of various makes. The qualitative study of desorbate IR-spectra has shown that they refer to the classes of arylaldehydes, diarylketones and carbonic acids. Under actual conditions first-circuit reactor coolant also has a wide set of products of its radiolysis, therefore the spectrum of coolant oxidaton products must be wider. It is noted that adsorption on silica gel, ASK of oxygen-bearing compounds which are present in ditolyl methane coolant has 2 stages

Structural analysis of the as-built IEA-R1 primary coolant piping system using a complete three dimensional model

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Martins, Lucas B.; Marcolin, Gabriel; Mattar Neto, Miguel

2011-01-01

IEA-R1 is an open pool type research reactor, moderated by light water and upgraded from 2 MW to 5 MW of operating power level. Heat generated in the reactor core is removed by a coolant system divided in two circuits, primary and secondary, composed by pumps, piping, heat exchangers, cooling tower, and some other auxiliary components. The 5 MW operating power level is now possible due to a modernization program started in 1996. As a part of the modernization program, ageing assessment studies recommend the replacement of one of the two heat exchangers in the circuit. To manage this replacement, modifications in the layout of the primary and secondary piping and supporting systems were performed, based on preliminary stress analysis study. Then, the aim of this work is to present the final stress analysis of the primary circuit. To reach this and taking the modifications of the primary into account, a 3D model of the whole circuit, in the as-built condition, was made. Stress results and discussions are shown. (author)

Engine Tune-Up Service. Unit 4: Secondary Circuit. Student Guide. Automotive Mechanics Curriculum.

Science.gov (United States)

Bacon, E. Miles

This student guide is for Unit 4, Secondary Circuit, in the Engine Tune-Up Service portion of the Automotive Mechanics Curriculum. It deals with how to test and service the secondary ignition circuit. A companion review exercise book and posttests are available separately as CE 031 215-216. An introduction tells how this unit fits into the total…

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

Computer programmes of the Power Research Institute for the analysis of processes in the primary coolant circuit and in the containment of a WWER plant in a loss-of-coolant accident

International Nuclear Information System (INIS)

Misak, J.

1976-01-01

A brief description is given of computer programmes for the analysis of loss-of-coolant accidents (LOCA) in WWER type reactors. The LENKA programme is intended for the thermal and hydraulic analysis of the consequences of such accidents in the primary coolant circuit. The SICHTA programme is intended for the detailed calculation of the time dependence of the axial and radial distribution of heat in fuel rods from steady-state to the flooding of the core. CHEMLOC is intended for the analysis of the heat history of the core and the extent of chemical reactions in LOCA when the emergency core cooling system is not operating. The TRACO I is intended for the analysis of the initial stage of the transient process in a full-pressure containment after LOCA (the computation of the time and spatial dependences of pressures and temperatures). TRACO III is intended for the computation of the long-term time dependence of pressure and temperature in the full-pressure containment after LOCA. (B.S.)

Reverse osmosis and its use at the nuclear power plants. Purification of primary circuit coolant by the means of reverse osmosis

International Nuclear Information System (INIS)

Kus, Pavel; Vonkova, Katerina; Kunesova, Katerina; Bartova, Sarka; Skala, Martin; Moucha, Tomáš

2014-01-01

This contribution is focused on the use of membrane technologies (e.g. reverse osmosis) for the primary coolant purification at the nuclear power plants. Currently, boric acid present in the primary coolant is preconcentrated at the evaporators, but their operation is very inefficient and expensive. Therefore, reverse osmosis was proposed as one of promising methods possibly replacing evaporators. The aim of the purification process is to achieve boric acid solution of a defined concentration (40 g/l) in the retentate stream in order to recycle it and reuse it in the primary circuit. Additionally, permeate flow should consist solely of pure water. To study the efficiency of several reverse osmosis modulus in the boric acid removal form the water solutions, experimental apparatus was constructed in our laboratory. It consists of the solution reservoir, pump and reverse osmosis modulus. The arrangement of experiments was batch and the retentate flow was refluxed to the feed solution. Several modulus of commercial reverse osmosis membranes were tested. The feed solution contained various concentrations of H 3 BO 3 , KOH, LiOH and NH 3 in order to simulate real primary coolant composition. Based on the experimental results, mathematical model was developed in order to optimize experimental conditions for the best results in primary coolant purification and boric acid preconcentration. (author)

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

Nuclear reactor of pressurized liquid coolant type

International Nuclear Information System (INIS)

Costes, D.

1976-01-01

The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

Engine Tune-up Service. Unit 4: Secondary Circuit. Review Exercise Book. Automotive Mechanics Curriculum.

Science.gov (United States)

Bacon, E. Miles

This book of pretests and review exercises is designed to accompany the Engine Tune-Up Service Student Guide for Unit 4, Secondary Circuit, available separately as CE 031 214. Focus of the exercises and pretests is testing and servicing the secondary ignition circuit. Pretests and performance checklists are provided for each of the seven…

Review of the OECD specialist meeting on continuous monitoring techniques for assuring coolant circuit integrity

International Nuclear Information System (INIS)

Thie, J.A.

1986-01-01

This article summarizes the OECD Specialist Meeting on Continuous Monitoring Techniques for Assuring Coolant Circuit Integrity held August 12-14, 1985, in London. The conference was organized by the Organization for Economic Cooperation and Development's (OECD's) Committee on the Safety for Nuclear Installations and hosted by Her Majesty's Nuclear Installation Inspectorate at King's College. Many other conferences have addressed analysis and inspection approaches to ensuring primary-system integrity, but the OECD meeting was structured to pay attention to the continuous monitoring approach - possibly the first conference to be so designed. The specific technologies represented were vibrations, noise (i.e., random fluctuations in signals), leaks, acoustic emission, and cyclic fatigue. Although water reactors dominate the papers, all reactor types were included. A diverse group of about 50 attendees from 11 countries participated, including representatives from utilities, suppliers, regulators, and researchers

Advances on the analysis of fast reactor core and coolant circuit structures

International Nuclear Information System (INIS)

Livolant, M.; Imazu, A.; Chang, Y.W.; Eggen, D.T.

1989-01-01

For the 10th SMiRT Conference, it has been decided to make general reviews of the accomplishments throughout the conferences. The aim of this paper is to make such a review in the field of fast reactor core and coolant circuit structures, which is now fully treated in division E. That was not true in the past: at the earliest conferences up to the 5th, the division E dealt with accidental studies among which the hypothetical core disruptive accident was the most important. So, to cover the subject from the first SMiRT to now, it has been necessary to search into all the past division in order to recover the studies fitting into the scope of the present division E. This has allowed a table showing the number of presented papers on the various topics at the SMiRT conferences to be set up (table I). Then, some significant topics have been studied in detail, highlighting the main accomplishments, but trying also to point out the shortcomings and the work still to be done, in view of the present state of art

5. International seminar on primary and secondary side water chemistry of nuclear power plants

International Nuclear Information System (INIS)

2001-01-01

The major subjects of the meetings are: water chemistry of primary and secondary coolant circuits of PWR type reactors (mainly WWER types), corrosion of steam generators, decontamination processes, treatment of radioactive waste waters and related subjects. All the 29 papers were individually indexed and abstracted for the INIS database. (R.P.)

5. International seminar on primary and secondary side water chemistry of nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-07-01

The major subjects of the meetings are: water chemistry of primary and secondary coolant circuits of PWR type reactors (mainly WWER types), corrosion of steam generators, decontamination processes, treatment of radioactive waste waters and related subjects. All the 29 papers were individually indexed and abstracted for the INIS database. (R.P.)

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

Operation results of the secondary circuits of the French PWR type power plant park

International Nuclear Information System (INIS)

Mercier, J.P.

1984-01-01

Global results of performances realized since 1981 by the French PWR 900 MW power plants (installed power, availability, casual or planned shutdowns); analysis of the behaviour (casual unavailability) comparing together the performances of the different components in the secondary circuit; behaviour of the principal materials of the secondary circuit and their weight in the unavailabilities of the whole French nuclear park [fr

Evaluation of CRUDTRAN code to predict transport of corrosion products and radioactivity in the PWR primary coolant system

International Nuclear Information System (INIS)

Lee, C.B.

2002-01-01

CRUDTRAN code is to predict transport of the corrosion products and their radio-activated nuclides such as cobalt-58 and cobalt-60 in the PWR primary coolant system. In CRUDTRAN code the PWR primary circuit is divided into three principal sections such as the core, the coolant and the steam generator. The main driving force for corrosion product transport in the PWR primary coolant comes from coolant temperature change throughout the system and a subsequent change in corrosion product solubility. As the coolant temperature changes around the PWR primary circuit, saturation status of the corrosion products in the coolant also changes such that under-saturation in steam generator and super-saturation in the core. CRUDTRAN code was evaluated by comparison with the results of the in-reactor loop tests simulating the PWR primary coolant system and PWR plant data. It showed that CRUDTRAN could predict variations of cobalt-58 and cobalt-60 radioactivity with time, plant cycle and coolant chemistry in the PWR plant. (author)

Modifications in secondary circuit chemistry of Atucha I nuclear power plant

International Nuclear Information System (INIS)

Iglesias, Alberto M.; Jimenez Rebagliatti, Raul; Gentilli, Nelida; Raffo Calderon, Maria C.; Gemini, Hector

1999-01-01

The CNA I secondary circuit presents, by design, some materials which are difficult to be compatible from the corrosion point of view. The presence of Cu alloys limits the use of ammonia (or products that for decomposition generates it) as pH regulator substance which would be convenient to minimize the corrosion processes. The pH limit value in agreement with the operative experience is 9.2. This value is below the one required to minimize the effects of corrosion on the carbon steel, which is present inside the secondary circuit with a considerable exposed area and under hydrodynamic and hydrothermal conditions that favor those processes. This corrosive effect diminishes below certain limits, i. d. if the pH value is increased. The realization of this study involves three stages at least: a)- Independent measurements and description of the circuit current state; b)- Laboratory experiences of the possible alternatives into replace NH3 as alkaline agent and to provide better control of the corrosion process, on Cu alloys as well as steel alloys; c)- Plant implementations of the actions that are feasible from the point of view of power station operation, in such a way that in the secondary circuit it minimizes the presence of ammonia in the vapor phase and at the same time, the possibility of increasing the pH of the liquid phase, to diminish the corrosion phenomena of carbon steel. (author)

Radionuclide activity in the secondary cooling circuit of the NPP with WWER-440

International Nuclear Information System (INIS)

Voronin, L.M.; Volkov, A.P.; Kozlov, V.F.; Luzanova, L.M.; Pashevich, V.I.

1979-01-01

On the basis of the Kol'skaya NPP operating experience the radioactivity of water and steam fed to turbines and boilers has been determined. The radioactivity of water in the primary circuit is demonstrated to be mainly due to 24 Na, 42 K, 18 F, sup(131-135)I, and in the secondary circuit - due to 131 I and 133 I. The data obtained testify to the permanent leakage of water from the primary circuit to the secondary one averaging 19+-12 g/hr at the pressure gradient between the circuits approximately 7.0 MPa. With the probability of 1.2x10 -1 -1.6x10 -2 (eff. reactor day) -1 or 1.2x10 -8 -1.6x10 -9 (kWxhr) -1 the leakage may for some time increase to 1 l/hr

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)

NPPs Secondary Circuit Piping Wall-Thinning Management in China

International Nuclear Information System (INIS)

Zhong Zhimin; Li Jinsong; Zheng Hui

2012-01-01

Since 1980s, secondary circuit piping wall-thinning incidents happened in nuclear power plants (NPPs) worldwide. Particularly Surry 2 and Mihama 3 accidents resulted from flow-accelerated corrosion (FAC), unplanned outage, huge fatalities and economic loss pushed whole industry to pay more attention on the wall-thinning problem.

Modeling a Printed Circuit Heat Exchanger with RELAP5-3D for the Next Generation Nuclear Plant

International Nuclear Information System (INIS)

2010-01-01

The main purpose of this report is to design a printed circuit heat exchanger (PCHE) for the Next Generation Nuclear Plant and carry out Loss of Coolant Accident (LOCA) simulation using RELAP5-3D. Helium was chosen as the coolant in the primary and secondary sides of the heat exchanger. The design of PCHE is critical for the LOCA simulations. For purposes of simplicity, a straight channel configuration was assumed. A parallel intermediate heat exchanger configuration was assumed for the RELAP5 model design. The RELAP5 modeling also required the semicircular channels in the heat exchanger to be mapped to rectangular channels. The initial RELAP5 run outputs steady state conditions which were then compared to the heat exchanger performance theory to ensure accurate design is being simulated. An exponential loss of pressure transient was simulated. This LOCA describes a loss of coolant pressure in the primary side over a 20 second time period. The results for the simulation indicate that heat is initially transferred from the primary loop to the secondary loop, but after the loss of pressure occurs, heat transfers from the secondary loop to the primary loop.

Variable cooling circuit for thermoelectric generator and engine and method of control

Science.gov (United States)

Prior, Gregory P

2012-10-30

An apparatus is provided that includes an engine, an exhaust system, and a thermoelectric generator (TEG) operatively connected to the exhaust system and configured to allow exhaust gas flow therethrough. A first radiator is operatively connected to the engine. An openable and closable engine valve is configured to open to permit coolant to circulate through the engine and the first radiator when coolant temperature is greater than a predetermined minimum coolant temperature. A first and a second valve are controllable to route cooling fluid from the TEG to the engine through coolant passages under a first set of operating conditions to establish a first cooling circuit, and from the TEG to a second radiator through at least some other coolant passages under a second set of operating conditions to establish a second cooling circuit. A method of controlling a cooling circuit is also provided.

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

Development and Simulation of Increased Generation on a Secondary Circuit of a Microgrid

Science.gov (United States)

Reyes, Karina

As fossil fuels are depleted and their environmental impacts remain, other sources of energy must be considered to generate power. Renewable sources, for example, are emerging to play a major role in this regard. In parallel, electric vehicle (EV) charging is evolving as a major load demand. To meet reliability and resiliency goals demanded by the electricity market, interest in microgrids are growing as a distributed energy resource (DER). In this thesis, the effects of intermittent renewable power generation and random EV charging on secondary microgrid circuits are analyzed in the presence of a controllable battery in order to characterize and better understand the dynamics associated with intermittent power production and random load demands in the context of the microgrid paradigm. For two reasons, a secondary circuit on the University of California, Irvine (UCI) Microgrid serves as the case study. First, the secondary circuit (UC-9) is heavily loaded and an integral component of a highly characterized and metered microgrid. Second, a unique "next-generation" distributed energy resource has been deployed at the end of the circuit that integrates photovoltaic power generation, battery storage, and EV charging. In order to analyze this system and evaluate the impact of the DER on the secondary circuit, a model was developed to provide a real-time load flow analysis. The research develops a power management system applicable to similarly integrated systems. The model is verified by metered data obtained from a network of high resolution electric meters and estimated load data for the buildings that have unknown demand. An increase in voltage is observed when the amount of photovoltaic power generation is increased. To mitigate this effect, a constant power factor is set. Should the real power change dramatically, the reactive power is changed to mitigate voltage fluctuations.

Experience running in and improvement of secondary circuit water chemistry of Kalinin NPP WWER-1000

International Nuclear Information System (INIS)

Noev, V.V.; Kukharev, N.D.; Otchenashev, G.D.; Guzeeva, G.I.; Kochetova, G.G.

1991-01-01

Basic characteristics of the secondary circuit water-chemical conditions at the Kalinin 1 and 2 reactors are presented. These are the WWER-1000 reactors with K-1000-1500 turbines. The analysis conducted makes it possible to conclude that all indicated values can meet the standards by introducing the hydrazine-ammonium regime in the secondary circuit with feedwater pH value equal to 9±0.2. Realization of design scheme for condensate-feeding circuit washing is necessary for acceleration of the water-chemical mode stabilization. Moreover the units should be equipped with automated chemical control instrumentation of a new generation

Testing methods of gaseous admixtures in HLMC circuits

International Nuclear Information System (INIS)

Shelemet'ev, V.M.; Martynov, P.N.; Askhadullin, R.Sh.; Storozhenko, A.N.; Sadovnichij, R.P.; Ivanov, I.I.

2014-01-01

Control of gas phase is the effective method for state diagnostics of circuit of nuclear power facilities with heavy liquid metal coolants. Use of developing in IPPE solid electrolyte and conductometric oxygen and hydrogen sensors, which are set directly in gas system of the primary circuit, allows to maintain continuously control of oxygen and hydrogen content as well as operational efficiency and accuracy of these parameters determination under various situations related with oxygen and hydrogen insertion into circuit. Sensors ensure long-term safe operation under extreme conditions of high temperatures, pressures, humidity, etc., and are advanced devices for application in nuclear power facilities with heavy liquid metal coolants [ru

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

Refurbishment of the IEAR1 primary coolant system piping supports

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2015-01-01

A partial replacement of the IEA-R1 piping system was concluded in 2014. This paper presents the study and the structural analysis of the IEA-R1 primary circuit piping supports, considering all the changes involved in the replacement. The IEA-R1 is a nuclear reactor for research purposes designed by Babcox-Willcox that is operated by IPEN since 1957. The reactor life management and modernization program is being conducted for the last two decades and already resulted in a series of changes, especially on the reactor coolant system. This set of components, divided in primary and secondary circuit, is responsible for the circulation of water into the core to remove heat. In the ageing management program that includes regular inspection, some degradation was observed in the primary piping system. As result, the renewing of the piping system was conducted in 2014. Moreover the poor condition of some original piping supports gave rise to the refurbishment of all piping supports. The aim of the present work is to review the design of the primary system piping supports taking into account the current conditions after the changes and refurbishment. (author)

Calculation methods and algorithms development of dynamic loadings on NPP secondary circuit equipment at shock and pulse actions

International Nuclear Information System (INIS)

Kuznetsov, D.V.; Kormilitsyn, V.M.; Proskuryakov, K.N.

2010-01-01

Calculation results of acoustic parameters fluctuations in low-pressure regenerative heating system of NPP with WWER-1000 type reactor were presented. The spectral structure of acoustic fluctuations was shown to depend on configuration of secondary circuit equipment, its geometrical sizes and operation mode. Estimations of natural oscillations frequencies of working medium pressure in the secondary circuit equipment were resulted. The developed calculation methods and algorithms are intended for revealing and prevention of initiation conditions of vibrations resonances in elements of the secondary circuit equipment with acoustic oscillations in working medium, both under operating conditions and in the design stage of the second circuit of NPP with WWER-1000 type reactor. Analysis of pass-band dependence on operation mode was carried out to solve the given problem [ru

Design analysis of a lead–lithium/supercritical CO2 Printed Circuit Heat Exchanger for primary power recovery

International Nuclear Information System (INIS)

Fernández, Iván; Sedano, Luis

2013-01-01

Highlights: • A design for a PbLi/CO 2 (SC) Printed Circuit Heat Exchanger which optimizes the pressure drop performance is proposed. • Numerical analyses have been performed to optimize the airfoil fins shape and arrangement. • SiC is proposed as structural material and tritium permeation barrier for the PCHE. • The integrated flux is larger than expected and allows reducing the CO 2 mass flow in this sector of the power cycle. • A transport model has been developed to evaluate the permeation of tritium from the liquid metal to the secondary CO 2 . -- Abstract: One of the key issues for fusion power plant technology is the efficient, reliable and safe recovery of the power extracted by the primary coolants. An interesting design option for power conversion cycles based on Dual Coolant Breeding Blankets (DCBB) is a Printed Circuit Heat Exchanger, which is supported by the advantages of its compactness, thermal effectiveness, high temperature and pressure capability and corrosion resistance. This work presents a design analysis of a silicon carbide Printed Circuit Heat Exchanger for lead–lithium/supercritical CO 2 at DEMO ranges (4× segmentation)

Nuclear reactor monitoring device

International Nuclear Information System (INIS)

Mihashi, Ishi; Honma, Hitoshi.

1993-01-01

The monitoring device of the present invention comprises a reactor core/reactor system data measuring and controlling device, a radioactivity concentration calculation device for activated coolants for calculating a radioactivity concentration of activated coolants in a main steam and reactor water by using an appropriate physical model, a radioactivity concentration correlation and comparison device for activated coolants for comparing correlationship with a radiation dose and an abnormality alarm device. Since radioactivity of activated primary coolants is monitored at each of positions in the reactor system and occurrence of leakage and the amount thereof from a primary circuit to a secondary circuit is monitored if the reactor has secondary circuit, integrity of the reactor system can be ensured and an abnormality can be detected rapidly. Further, radioactivity concentration of activated primary circuit coolants, represented by 16 N or 15 C, is always monitored at each of positions of PWR primary circuits. When a heat transfer pipe is ruptured in a steam generator, leakage of primary circuit coolants is detected rapidly, as well as the amount of the leakage can be informed. (N.H.)

The propagation of pressure pulsations in the primary circuit of power plant A1

International Nuclear Information System (INIS)

Pecinka, L.

1976-01-01

A classification is made of the exciting forces of pressure pulsations in the primary coolant circuit with forced coolant circulation. A mathematical model is constructed of the propagation of pressure pulsations in the system and examples of measurements are given. The measurement methods used and the methods for the generalization of obtained data are assessed. The methods and results of the measurements of hydrodynamic pressure pulsations in a closed primary circuit with forced coolant circulation of the A-1 nuclear power plant are given. (F.M.)

Simulation of the dynamic behaviour of the secondary circuit of a WWER-440 type nuclear power plant Pt. 1

International Nuclear Information System (INIS)

Gacs, A.; Janosy, J.S.; Kiss, Zs.

1987-07-01

This report describes the simulation model of the secondary circuit of a WWER-440 type nuclear power plant. The goal of this modelling is to simulate normal and small abnormal transients in a Basic Principles Simulator. The earlier reports describing the dynamic simulation of primary circuit of a WWER-440 nuclear power plant are KFKI--1983-127 and KFKI--1985-08. At present the controllers of the secondary circuit are not simulated. Finally, some simulation results are presented. (author)

Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

International Nuclear Information System (INIS)

Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

2013-01-01

A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

Loose part trapping device in the secondary circuit of a steam generator

International Nuclear Information System (INIS)

Poussin, C.; Ayme, H.

1993-01-01

The trap is principally formed from mesh filter placed across the annular section of the secondary circuit. The maximum mesh size is less than the distance separating the tubes so that the loose parts cannot caught between the tubular bundle tubes. 6 figs

Applied model of through-wall crack of coolant vessels of WWER-type reactors

International Nuclear Information System (INIS)

Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

2010-01-01

We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

Simplified model of a PWR primary coolant circuit

International Nuclear Information System (INIS)

Souza, A.L. de; Faya, A.J.G.

1988-01-01

The computer program RENUR was developed to perform a very simplified simulation of a typical PWR primary circuit. The program has mathematical models for the thermal-hydraulics of the reactor core and the pressurizer, the rest of the circuit being treated as a single volume. Heat conduction in the fuel rod is analysed by a nodal model. Average and hot channels are treated so that the bulk response of the core and DNBR can be evaluated. A Homogenenous model is employed in the pressurizer. Results are presented for a steady-state situation as well as for a loss of load transient. Agreement with the results of more elaborate computer codes is good with substantial reduction in computer costs. (author) [pt

Thermal hydraulic tradeoffs in the design of IRIS primary circuit

International Nuclear Information System (INIS)

Oriani, L.; Lombardi, C.; Ricotti, M.E.; Paramonov, D.; Carelli, M.; Conway, L.

2001-01-01

IRIS (International Reactor Innovative and Secure) is currently being developed by an international consortium, led by Westinghouse and including universities. In order to achieve high level of safety, reduce complexity and capital cost, and enhance proliferation resistance, an integral primary circuit configuration has been selected. The integral configuration (the core, steam generators, coolant pumps, pressurizer and control rods are all contained within the reactor vessel) has no loop piping and thereby eliminates the possibility of large loss of coolant accidents. If the reactor vessel and components are designed for a very high level of natural circulation, which is promoted by an integral design, the consequence of loss of flow accidents can be significantly reduced or even completely eliminated. Core and integral primary circuit design optimization has been performed using the OSCAR computer code, a specialized tool for the analyses of the IRIS primary system developed at POLIMI. Results of trade-off studies of various in-vessel configurations explored to achieve tight packaging and high serviceability and/or replacement of components such as steam generators and pumps are reported. Effects of changes in secondary side parameters and steam generator design on system efficiency were explored together with the optimization of the vessel and steam generator dimensions and costs. The aim of the trade-off analyses was to limit the design space, and select a reference configuration for the IRIS reactor. (author)

The experimental definition of the acoustic standing wave series shapes, formed in the coolant of the primary circuit of VVER-440 type reactor

International Nuclear Information System (INIS)

Bulavin, V.V.; Pavelko, V.I.

1995-01-01

On the basis of pressure fluctuation measurements in some primary circuit loops at 2 nd Unit of Kola NPP with VVER-440 type reactors, the shapes of acoustic standing waves (ASW) were determined at frequencies corresponding to four minimal oscillation eigenfrequencies in the primary circuit coolant. On identification of the ASW modes and properties, experimental results based on six circulating loops in symmetric arrangement allowed determination of the three-dimensional space structure of the wave nodes and antinodes inside and outside of the reactor vessel (RV). As part of this analysis, the geometric features of the primary circuit that caused the formation of these standing waves were identified. Differences in each ASW shape were shown to cause different individual effects on the neutron field in the reactor core and on fuel assembly vibration. This has been partially confirmed by ex-core neutron ionization chamber noise analysis. One type of ASW, possessing an antinode inside the RV, can be used for measurement of the pressure coefficient of reactivity. However, this must be done with care to avoid the potential for incorrect results in some cases. The results presented in this paper can be readily extended to other VVER type reactors with both odd and even number of loops. (author)

Influence of the secondary circuit chemistry on the characteristics of the filmed carbon steel components

International Nuclear Information System (INIS)

Radulescu, M.; Pirvan, I.; Dinu, A.; Fulger, M.; Lucan, D.; Anghel, C.

2001-01-01

Some correlations are established between the characteristics of the superficial films formed on the carbon steel components in some operation conditions of a secondary circuit from Nuclear Power Station (NPS), and the specific parameters of the aqueous environment in which these films were formed. The main parameters studied, specific to a secondary circuit environment, are: the value of pH and the substances used to regulate it and, respectively, the oxygen concentration dissolved in the aqueous environment. These filmed samples by autoclaving were studied by: X rays diffraction method; metallographic microscopy and electrochemical methods such as: potentiodynamic and electrochemical impedance spectroscopy (E.I.S.). (R.P.)

Technical specifications: McGuire Nuclear Station, Unit Nos. 1 and 2. Docket Nos. 50-369 and 50-370

International Nuclear Information System (INIS)

Anderson, F.

1983-03-01

Information is presented concerning safety limits for operation; control systems; power distribution limits; instrumentation; primary coolant circuit; emergency core cooling systems; containment systems; steam turbines and secondary coolant circuits; electrical equipment; refueling operations; radioactive effluents; and reactor site

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

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

RF DEMO ceramic helium cooled blanket, coolant and energy transformation systems

International Nuclear Information System (INIS)

Kovalenko, V.; Leshukov, A.; Poliksha, V.; Popov, A.; Strebkov, Yu.; Borisov, A.; Shatalov, G.; Demidov, V.; Kapyshev, V.

2004-01-01

RF DEMO-S reactor is a prototype of commercial fusion reactors for further generation. A blanket is the main element unit of the reactor design. The segment structure is the basis of the ceramic blanket. The segments mounting/dismounting operations are carried out through the vacuum vessel vertical port. The inboard/outboard blanket segment is the modules welded design, which are welded by back plate. The module contains the back plate, the first wall, lateral walls and breeding zone. The 9CrMoVNb steel is used as structural material. The module internal space formed by the first wall, lateral walls and back plate is used for breeding zone arrangement. The breeding zone design based upon the poloidal BIT (Breeder Inside Tube) concept. The beryllium is used as multiplier material and the lithium orthosilicate is used as breeder material. The helium at 0.1 MPa is used as purge gas. The cooling is provided by helium at 10 MPa. The coolant supply/return to the blanket modules are carrying out on the two independent circuits. The performed investigations of possible transformation schemes of DEMO-S blanket heat power into the electricity allowed to make a conclusion about the preferable using of traditional steam-turbine facility in the secondary circuit. (author)

On steady-state concentrations of ammonia and molecular hydrogen in the primary circuit of the WWER-1000 reactors

International Nuclear Information System (INIS)

Arkhipov, O.P.; Bugaenko, V.L.; Kamakchi, S.A.

1997-01-01

It is shown that the MORAVA-N2 software package describes well the coolant state in the primary circuit of an actual reactor facility with the WWER-1000 during on-load operation. It permits using the package for analysis of process perturbation effect on the coolant composition. Specific feature of ammonia radiation chemistry in the primary circuit of a reactor facility with the WWER-1000, assuring the rates hydrogen concentration in the coolant with ammonia concentration variation in the coolant within wide limits, when reactor operates on power, can be mentioned by way of example, the fact being ascertained in this study

Analysis of Consequences in the Loss-of-Coolant Accident in Wendelstein 7-X Experimental Nuclear Fusion Facility

Energy Technology Data Exchange (ETDEWEB)

Uspuras, E., E-mail: algis@mail.lei.lt [Laboratory of Nuclear Installations Safety, Lithuanian Energy Institute, Kaunas (Lithuania)

2012-09-15

Full text: Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Starting 2007, Lithuanian energy institute (LEI) is a member of European Fusion Development Agreement (EFDA) organization. LEI is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) in the frames of EFDA project by performing safety analysis of fusion device W7-X. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. The W7-X facility divertor cooling system consists of two coolant circuits: the main cooling circuit and the so-called 'baking' circuit. Before plasma operation, the divertor and other invessel components must be heated up in order to 'clean' the surfaces by thermal desorption and the subsequent pumping out of the released volatile molecules. The rupture of pipe, providing water for the divertor targets during the 'baking' regime is one of the critical failure events, since primary and secondary steam production leads to a rapid increase of the inner pressure in the plasma (vacuum) vessel. Such initiating event could lead to the loss of vacuum condition up to overpressure of the plasma vessel, damage of in-vessel components and bellows of the ports. In this paper the safety analysis of 40 mm inner diameter coolant pipe rupture in cooling circuit and discharge of steam-water mixture through the leak into plasma vessel during the W7-X no-plasma 'baking' operation mode is presented. For the analysis the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers) and plasma vessel was developed by employing system thermal-hydraulic state-of-the-art RELAP5 Mod 3.3 code. This paper demonstrated, that the developed RELAP5 model allows to analyze the processes in divertor cooling system and plasma vessel

Thermal circuit and supercritical steam generator of the BGR-300 nuclear power plant

International Nuclear Information System (INIS)

Afanas'ev, B.P.; Godik, I.B.; Komarov, N.F.; Kurochnkin, Yu.P.

1979-01-01

Secondary coolant circuit and a steam generator for supercritical steam parameters of the BGR-300 reactor plant are described. The BGR-300 plant with a 300 MW(e) high-temperature gas-cooled fast reactor is developed as a pilot commercial plant. It is shown that the use of a supercritical pressure steam increases the thermal efficiency of the plant and descreases thermal releases to the environment, permits to use home-made commercial turbine plants of large unit power. The proposed supercritical pressure steam generator has considerable advantages from the viewpoint of heat transfer and hydrodynamical processes

Corrosion Behavior of Carbon Steel Coated with Octadecylamine in the Secondary Circuit of a Pressurized Water Reactor

Science.gov (United States)

Jäppinen, Essi; Ikäläinen, Tiina; Järvimäki, Sari; Saario, Timo; Sipilä, Konsta; Bojinov, Martin

2017-12-01

Corrosion and particle deposition in the secondary circuits of pressurized water reactors can be mitigated by alternative water chemistries featuring film-forming amines. In the present work, the corrosion of carbon steel in secondary side water with or without octadecylamine (ODA) is studied by in situ electrochemical impedance spectroscopy, combined with weight loss/gain measurements, scanning electron microscopy and glow-discharge optical emission spectroscopy. The impedance spectra are interpreted using the mixed-conduction model to extract kinetic parameters of oxide growth and metal dissolution through it. From the experimental results, it can be concluded that ODA addition reduces the corrosion rate of both fresh and pre-oxidized carbon steel in secondary circuit significantly by slowing down both interfacial reactions and transport through the oxide layer.

Radionuclide deposits on heat transfer surfaces in a circumt with dissociating N2O4 coolant

International Nuclear Information System (INIS)

Dolgov, V.M.; Katanaev, A.O.; Komissarov, F.D.

1984-01-01

Radionuclides deposits on heat transfer surfaces of a circuit with dissociating coolant are studied. The areas of preferential deposition of 54 Mn, 51 Cr, 134 Cs and their distribution along the heating and cooling surfaces are determined. The comparison of the obtained data on the nuclide and chemical compositions of the deposits in the areas of N 2 O 4 coolant heating and cooling shows that 54 Mn, 51 Cr, 134 Cs deposit preferentially on heat transfer surfaces in the area of the coolant heating. Fixed and movable deposits consists of the structural material oxides. The quantity of radionuclides in the deposits on the surfaces of heat transfer tubes in the area of cooling decreases with the coolant temperature drop

Integral forged pump casing for the primary coolant circuit of a nuclear reactor: Development in design, forging technology, and material

International Nuclear Information System (INIS)

Austel, W.; Korbe, H.

1986-01-01

Developments in the forging of large casings for primary circuit coolant pumps for light water reactors in Germany are demonstrated beginning with the multiple forging fabricated version and ending with the integral forged type. This version is the result of the joint efforts of the pump manufacturer and the forgemaster after a cost-gain evaluation and represents an optimum solution in view of its functional and economical performance and also considering the high requirements for mechanical-technological properties, including homogeneity of the material. The development from 22 NiMoCr 3 7/A 508 Class 2 to 20 MnMoNi 5 5/A 508 Class 3 and their optimization will be demonstrated. This development is based mainly on minimizing the sulfur content and on vacuum carbon deoxidation (VCD), which results in a reduction of the A-segregations, in improving fracture toughness and isotropy, and in the desired fine-grain structure

Installations having pressurised fluid circuits

International Nuclear Information System (INIS)

Rigg, S.; Grant, J.

1977-01-01

Reference is made to nuclear installations having pressurised coolant flow circuits. Breaches in such circuits may quickly result in much damage to the plant. Devices such as non-return valves, orifice plates, and automatically operated shut-off valves have been provided to prevent or reduce fluid flow through a breached pipe line, but such devices have several disadvantages; they may present large restrictions to normal flow of coolant, and may depend on the operation of ancillary equipment, with consequent delay in bringing them into operation in an emergency. Other expedients that have been adopted to prevent or reduce reverse flow through an upstream breach comprise various forms of hydraulic counter flow brakes. The arrangement described has at least one variable fluid brake comprising a fluidic device connected into a duct in the pressurised circuit, the device having an inlet, an outlet, a vortex chamber between the inlet and outlet, a control jet for introducing fluid into the vortex chamber, connections communicating the inlet and the outlet into one part of the circuit and the control jet into another region at a complementary pressure so that, in the event of a breach in the circuit in one region, fluid passes from the other region to enter the vortex chamber to stimulate pressure to create a flow restricting vortex in the chamber that reduces flow through the breach. The system finds particular application to stream generating pressure tube reactors, such as the steam generating heavy water reactor at UKAEA, Winfrith. (U.K.)

Radiolytic reactions in the coolant of helium cooled reactors

International Nuclear Information System (INIS)

Tingey, G.L.; Morgan, W.C.

1975-01-01

The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

Method of nuclear power plant transition to partial loa

International Nuclear Information System (INIS)

Ilyunin, V.G.; Kuznetsov, I.A.; Murogov, V.M.; Shmelev, A.N.

1977-01-01

Five variants are proposed of a nuclear power plant with two reactors, of which one a breeder reactor, suitable for work under partial electrical load. Both reactors have a common secondary coolant circuit with a turbine, the breeder reactor operates at a lower coolant temperature in the primary coolant circuit than the other. In the transition of the power plant to the partial electrical load mode, the coolants of both reactors mix at a certain ratio. Heating of the turbine working medium is controlled so that the breeder reactor output increases rather than decreases. The breeding ratio is thereby also increased. Excess heat in the breeder reactor primary circuit is removed via a separate coolant circuit provided with a heat consumer. (Z.M.)

Helium impurities in a PNP-primary coolant circuit

International Nuclear Information System (INIS)

Reif, M.

1981-01-01

The concentration of impurities to be expected have been defined in consideration of recent findings concerning the rates of infiltration and formation and the reaction mechanisms of the impurity components in the circuit. The data obtained correspond with the requirements on the metallic high-temperature components as well as with the requirements of limited graphite corrosion. (DG) [de

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)

Full Scope Modeling and Analysis on the Secondary Circuit of Chinese Large-Capacity Advanced PWR Based on RELAP5 Code

Directory of Open Access Journals (Sweden)

Dao-gang Lu

2015-01-01

Full Text Available Chinese large-capacity advanced PWR under construction in China is a new and indispensable reactor type in the developing process of NPP fields. At the same time of NPP construction, accident sequences prediction and operators training are in progress. Since there are some possible events such as feedwater pumps trip in secondary circuit may lead to severe accident in NPP, training simulators and engineering simulators of CI are necessary. And, with an increasing proportion of nuclear power in China, NPP will participate in regulating peak load in power network, which requires accuracy calculation and control of secondary circuit. In order to achieve real-time and full scope simulation in the power change transient and accident scenarios, RELAP5/MOD 3.4 code has been adopted to model the secondary circuit for its advantage of high calculation accuracy. This paper describes the model of steady state and turbine load transient from 100% to 40% of secondary circuit using RELAP5 and provides a reasonable equivalent method to solve the calculation divergence problem caused by dramatic two-phase condition change while guaranteeing the heat transfer efficiency. The validation of the parameters shows that all the errors between the calculation values and design values are reasonable and acceptable.

Parametric study of the primary and secondary systems of the CAREM-25 reactor on steady state

International Nuclear Information System (INIS)

Halpert, Silvia; Vazquez, Luis

2000-01-01

In the CAREM-25 reactor the primary coolant flows by natural convection that's why the flow is established when the balance between the buoyancy force and friction pressure drop through circuit is obtained. This paper presents a parametric study on primary and secondary systems of the reactor on steady state, for different values of some thermohydraulics parameters: safety factor on friction loss pressure calculations (f), steam generator heat transfer area (A T ) and primary pressure (P P ). The ESCAREM 2.08 thermohydraulic code, which calculates the primary system behavior for steady state conditions, was used for this study. The conclusions of this study are: -) There was a variation of the 15% on the primary coolant flow when the safety factor was changed a 50 %; -) The primary and secondary systems conditions do not change when the power is less than 100 MW; -) Between 100 and 110 MW the decrease of the heat transfer area produces an important change on the secondary systems conditions: the outlet steam generator temperature decrease and there is an important rice in the flow; -) The primary pressure could decrease up to 11.4 MPa without violating turbine requirements. (author)

SODEXPERT: help to PWR plant management to prevent secondary circuit corrosion

International Nuclear Information System (INIS)

Eon-Duval, P.; Fiquet, J.M.; Langlet, J.P.

1996-01-01

Since about 10 years, problems of secondary circuit corrosion have raised for PWR plant management. The watch staff can't be asked the physicochemical knowledge requested for a proper interpretation of the various probes outputs. So an expert-system has been performed to help the identification of dangerous situation from a corrosion point of view, and immediately start the PWR managing action. This software has been successfully tested and validated. (D.L.). 5 figs., 4 photos

Using of new chemical regime on secondary circuit in nuclear power plant 'Kozloduy' - Bulgaria

International Nuclear Information System (INIS)

Minkova, K.; Stanchev, S.; Kalpakchiev, Z.

2010-01-01

There are two WWER-1000 - Russian design units (№5 and №6) working in the Bulgarian Nuclear Power Plant 'Kozloduy', which are equipped with horizontal steam generators. The report describes the chronology of the stages in the improvement of the water chemistry regime on the secondary circuit at both of the units 5 and 6 from their start up till now. As a final result of our purposeful activity is done a full exchange of copper alloys with stainless steel in regenerative LP heat exchangers and condensers in the secondary circuit equipment. That way we created preconditions for cardinal change and improvement of the water chemistry regime of the circuit. For that purpose is realized a program for optimization in maximal degree of the water chemistry regime of the secondary circuit of the WWER-1000 Units. The program includes the preliminary study, the stage of experimental operation and the stage of establishing the new regime into regular exploitation. The preliminary study included the definition of the coefficients of distribution of the ammonia in a real operation conditions in the secondary circuit and rating the expected results from the applying of different varieties of the secondary circuit water chemistry. It is taken into account the balance plans for distribution of the corrective reagents, the expected pH values at working temperature (pHt), the transport of products of corrosion to steam generators and the abilities of the installation for purification of the blow down water. As a final result a mixed ammonia-ethanolamine water chemistry regime has been chosen, characterized with pH 25 =9.9(9.85-9.95). The special feature of this regime is that, both the corrective reagents ethanolamine and ammonia are dosed in the feed up water as a mixed solution with permanent proportion of the concentrations. This water chemistry treatment combines the advantages of high ammonia AVT water chemistry related to specific constructive steels in low

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

Characterization of oxide film layers formed on A106 B carbon steel in simulated secondary coolant conditions of nuclear power plants

International Nuclear Information System (INIS)

Strack, M.; Bordoni, R.; Chocron, M.; Olmedo, A.M.; Zampieri, G.

2011-01-01

The water chemistry of the secondary coolant in the majority Nuclear Power Plants is controlled by AVT (All Volatile Treatment) procedure, wherein volatile amines are use to maintain the alkaline pH required for minimizing the corrosion of structural materials which one of them is Carbon Steel. In this procedure: hydrazine, morpholine and ethanolamine are used commonly as conditioning reagents. In this context, experiments were carried out by exposing carbon steel A106 B samples in a simulated secondary coolant in order to study the nature of the oxide films. The tests were performed in a static autoclave at 260 ºC using two media: 1) Hydrazine + morpholine and 2) Hydrazine + ethanolamine during different exposure periods up to ≈1020 h. The composition, surface morphology, X-ray diffraction, a chemical descaling procedure were used- XPS, was also employed, to analyze the films grown during ≈1020 h in both media. The characterization showed that magnetite was the main corrosion product formed in the films grown in the two media. The material weight loss (W) could be fitted by a law of the type W = k t n , up to 1020 h of exposure, resulting in n =0,42, k = 6,24 for films grown in medium 1) and n = 0,39, k =6,08 for films grown in medium 2); where W is in mg/d m 2 and t in h. (author) [es

The effect of the PWR secondary circuit water chemistry on erosion corrosion

International Nuclear Information System (INIS)

Kaplan, J.

1993-07-01

The secondary circuit of WWER-440 and WWER-1000 reactors is described. The causes of erosion corrosion are outlined, and the effects of the physical properties and chemical composition of water are discussed with emphasis on specific conductivity and concentrations of oxygen, ammonia, iron, sodium, silicon and organics. Described are corrective actions to eliminate the deviations from the normal state during reactor power reduction or reactor shutdown. (J.B.)

Components of the primary circuit of LWRs

International Nuclear Information System (INIS)

1980-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

Analytical and sampling problems in primary coolant circuits of PWR-type reactors

International Nuclear Information System (INIS)

Illy, H.

1980-10-01

Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

Evaluation of specific activity in the primary circuit of SMART-P

International Nuclear Information System (INIS)

Kim, Ah Young; Choi, Byung Seon; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

2005-01-01

SMART-P is a soluble boron free reactor, and the ammonia is used as a pH reagent. The titanium alloy, which has a high corrosion resistance, is chosen as a steam generator tube material. Despite these design features to achieve the corrosion reduction, it is expected that SMART-P exhibits a relatively high specific activity in the coolant due to the lack of purification during the power operation. The main reason for the high specific activity is the activation and transportation of the corrosion products that released from the primary circuit surfaces. The objective of this work is to analyze the corrosion product activity in the primary circuit of SMART-P using a multi-region model, KORA. This model, which is incorporated with the mass and activity transport between the dissolved corrosion products in the coolant and the surface, describes the specific activity of corrosion products in coolant and on the surfaces according to the operation modes

Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

OpenAIRE

Lefèvre, Grégory; Živković, Ljiljana S.; Jaubertie, Anne

2012-01-01

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspec...

Evaluation of steam as a potential coolant for nonbreeding blanket designs

International Nuclear Information System (INIS)

Stevens, H.C.; Misra, B.; Youngdahl, C.K.

1978-01-01

A steam-cooled nonbreeding blanket design has been developed as an evolution of the Argonne Experimental Power Reactor (EPR) studies. This blanket concept complete with maintenance considerations is to function at temperatures up to 650 0 C utilizing nickel-based alloys such as Inconel 625. Thermo-mechanical analyses were carried out in conjunction with thermal hydraulic analysis to determine coolant chennel arrangements that permit delivery of superheated steam at 500 0 C directly to a modern fossil plant-type turbine. A dual-cycle system combining a pressurized water circuit coupled with a superheated steam circuit can produce turbine plant conversion efficiencies approaching 41.5%

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)

Calorimetric and reactor coolant system flow uncertainty

International Nuclear Information System (INIS)

Bates, L.; McLean, T.

1991-01-01

This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

Nuclear power plants

International Nuclear Information System (INIS)

Kiyokawa, Teruyuki; Soman, Yoshindo.

1985-01-01

Purpose: To constitute a heat exchanger as one unit by integrating primary and secondary coolant circuits with secondary coolant circuit and steam circuit into a single primary circuit and steam circuit. Constitution: A nuclear power plant comprises a nuclear reactor vessel, primary coolant pipeways and a leakage detection system, in which a dual-pipe type heat exchanger is connected to the primary circuit pipeway. The heat conduction tube of the heat exchanger has a dual pipe structure, in which the inside of the inner tube is connected to the primary circuit pipeway, the outside of the outer tube is connected to steam circuit pipeway and a fluid channel is disposed between the inner and outer tubes and the fluid channel is connected to the inside of an expansion tank for intermediate heat medium. The leak detection system is disposed to the intermediate heat medium expansion tank. Sodium as the intermediate heat medium is introduced from the intermediate portion (between the inner and outer tubes) by way of inermediate heat medium pipeways to the intermediate heat medium expansion tank and, further, to the intermediate portion for recycling. (Kawakami, Y.)

Draft of diagnostic techniques for primary coolant circuit facilities using control computer

International Nuclear Information System (INIS)

Suchy, R.; Procka, V.; Murin, V.; Rybarova, D.

A method is proposed of in-service on-line diagnostics of primary circuit selected parts by means of a control computer. Computer processing will involve the measurements of neutron flux, pressure difference in pumps and in the core, and the vibrations of primary circuit mechanical parts. (H.S.)

Components of the LWR primary circuit. Pt. 2

International Nuclear Information System (INIS)

1984-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 0 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

Turning into carbonate the residual sodium left in BN-350 circuits may alleviate concerns over their long term safe confinement

International Nuclear Information System (INIS)

Rahmani, L

2000-01-01

After the coolant is drained from the reactor vessel and from the primary and secondary circuits of the BN-350 nuclear power plant, what sodium is left in ponds and films may amount to hundreds of kilograms. For the long term safe storage period which is to follow, preliminary safety analyses (e.g. derived from those made for French sodium cooled reactors) might show that the risks incurred through loss of leaktightness are significant. The ingress of moisture into the circuits would generate, by reaction with the sodium, two undesirable products : sodium hydroxide and hydrogene. Even when considering that water would enter the circuits progressively, so that the heat of the reaction does not give rise to over-pressure, some main risk factors remain. The most promising solution to this challenge appears to be the carbonation of the sodium residues, by progressive diffusion of an appropriate association of carbon dioxyde and water vapour through the inert gaseous medium which fills the circuits. The desired product is porous sodium hydrogenocarbonate

Radioisotopes in the primary circuit of a fast reactor

International Nuclear Information System (INIS)

Berlin, M.; Cauvin, M.

1976-01-01

In the frame of the research performed to understand the behaviour of the radioactive isotopes of iodine in the primary coolant circuit of fast reactor, a simple theoretical model is proposed. Results concerning PHENIX and RAPSODIE are given

Optimization of a primary circuit of the nuclear power plant from the vibration point of view

International Nuclear Information System (INIS)

Dupal, J.; Zeman, V.

2003-01-01

The primary circuit of the nuclear power plant (NPP) as a dynamical vibrating system can be disturbed by various excitation including earthquake or pressure pulsation generated by main circulation pumps (MCP). Especially, unpleasant pulsation vibration growth can be caused by the small differences of revolutions between main circulation pumps of the individual coolant loops. This growth corresponds to the well known beats. The paper deals with an approach to the improving and optimization of dynamical properties of the whole primary circuit system including the reactor and coolant loops under pressure pulsation. (author)

The operating reliability of the reactor coolant pump

International Nuclear Information System (INIS)

Grancy, W.

1996-01-01

There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

Method for investigation of various iodine species in the primary coolant of the nuclear power plant in Paks

International Nuclear Information System (INIS)

Volent, G.; Gimesi, O.; Solymosi, J.

1996-01-01

Iodine isotopes formed in the course of fission in nuclear reactors may be present in the primary coolant in different oxidation states, i.e., in different chemical forms. It is important to know the chemical forms and their proportions in order to asses the environmental effect of the emitted iodine and the performance of air filters used in the primary circuit for binding iodine, species, since both depend on the chemical forms in which it is present. Volatile components were separated from water samples taken separately from each block of the nuclear power station by purging with inert gas, then the aerosol, iodine vapour and alkyl iodides were selectively bound on the filter system of the 'KOMBI' sampler. I 3 - , I - , IO - , IO 3 - and IO 4 - left in the aqueous phase after purging were separated by consecutive physical and chemical procedures (extraction, isotope exchange, reduction). The results of the investigations have shown that the water technology used in the Nuclear Power Plant in Paks is appropriate with respect to the radioiodine balance. Iodine was found to be predominant species, and no volatile iodine species were found to be present in the primary coolant. Volatile iodine species sometimes appearing in emissions may be formed from leaching waters due to secondary effects. (author)

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.)

Responses to Small Break Loss of Coolant Accidents for SMART

International Nuclear Information System (INIS)

Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

2004-01-01

The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

2003-01-01

To support the development of a Probabilistic Safety Assessment (PSA) model usable in Riskinformed Applications (RIA) for Korea Standard Nuclear power Plants (KSNP), we have performed a thermal hydraulic analysis of Aggressive Secondary Cooldown (ASC) in a 2-inch Small Break Loss Of Coolant Accident (SBLOCA) with a total loss of High Pressure Safety Injection (HPSI). The present study focuses on the estimation of the success criteria of ASC, and the enhanced understanding of the detailed thermal hydraulic behavior and phenomena. The results have shown that the Reactor Coolant System (RCS) pressure can be reduced to the Low Pressure Safety Injection (LPSI) operation conditions without core damage. It was also shown that more relaxed success criteria compared to those in the previous PSA models of KSNP could be used in the new PSA model. However, it was found that the results could be affected by various parameters related with ASC operation, i.e., reference temperature for the calculation of the cooldown rate and its control method

Modeling the transport of nitrogen in an NPP-2006 reactor circuit

Science.gov (United States)

Stepanov, O. E.; Galkin, I. Yu.; Sledkov, R. M.; Melekh, S. S.; Strebnev, N. A.

2016-07-01

Efficient radiation protection of the public and personnel requires detecting an accident-initiating event quickly. Specifically, if a heat-exchange tube in a steam generator is ruptured, the 16N radioactive nitrogen isotope, which contributes to a sharp increase in the steam activity before the turbine, may serve as the signaling component. This isotope is produced in the core coolant and is transported along the circulation circuit. The aim of the present study was to model the transport of 16N in the primary and the secondary circuits of a VVER-1000 reactor facility (RF) under nominal operation conditions. KORSAR/GP and RELAP5/Mod.3.2 codes were used to perform the calculations. Computational models incorporating the major components of the primary and the secondary circuits of an NPP-2006 RF were constructed. These computational models were subjected to cross-verification, and the calculation results were compared to the experimental data on the distribution of the void fraction over the steam generator height. The models were proven to be valid. It was found that the time of nitrogen transport from the core to the heat-exchange tube leak was no longer than 1 s under RF operation at a power level of 100% N nom with all primary circuit pumps activated. The time of nitrogen transport from the leak to the γ-radiation detection unit under the same operating conditions was no longer than 9 s, and the nitrogen concentration in steam was no less than 1.4% (by mass) of its concentration at the reactor outlet. These values were obtained using conservative approaches to estimating the leak flow and the transport time, but the radioactive decay of nitrogen was not taken into account. Further research concerned with the calculation of thermohydraulic processes should be focused on modeling the transport of nitrogen under RF operation with some primary circuit pumps deactivated.

Fast reactors bulk sodium coolant disposal NOAH process application

International Nuclear Information System (INIS)

Magny, E. de; Berte, M.

1997-01-01

Within the frame of the fast reactors decommissioning, the becoming of contaminated sodium coolant from primary, secondary and auxiliary circuits is an important aspect. The 'NOAH' sodium disposal process, developed by the French Atomic Energy Commission (CEA), is presented as the only process, for destroying large quantities of contaminated sodium, that has attained industrial status. The principles and technical options of the process are described and main advantages such as safety , operating simplicity and compactness of the plant are put forward. The process has been industrially validated in 1993/1994 by successfully reacting the 37 metric tons of primary contaminated sodium from the French Rapsodie experimental reactor. The main outstanding aspects and experience gained from this so called 'DESORA' operation (DEstruction of SOdium from RApsodie) are recalled. Another industrial application concerns the current project for destroying more than 1500 metric tons of contaminated sodium from the British PFR (Prototype Fast Reactor) in Scotland. Although the design is in the continuity of DESORA, it has taken into account the specific requirements of PFR application and the experience feed back from Rapsodie. The main technical options and performances of the PFR sodium reaction unit are presented while mentioning the design evolution. (author)

BWR type nuclear power plant

International Nuclear Information System (INIS)

Nei, Hiromichi; Hashiguchi, Isao; Inai, Nobuhiko.

1996-01-01

A heat exchanger is disposed between a reactor pressure vessel and a turbine, an inlet of a primary circuit of the heat exchanger is connected to a steam pipeline, an exit of the primary circuit of the heat exchanger is connected to a primary coolant pipeline, the primary coolant pipeline is connected to a feed water pipeline, a secondary circuit steam pipeline connected to the heat exchanger is connected to the turbine and a condensate circuit from the turbine is connected to the secondary coolant pipeline connected to the heat exchanger. Steams generated in the reactor are once flown into the heat exchanger to heat secondary coolants indirectly in the heat exchanger, and the generated steams are introduced to the steam turbine. Incondensible gases generated from the reactor and inflowing to the primary side of the heat exchanger are introduced, together with a portion of the steams, to a small-sized condensator passing through steam pipelines in the vicinity of a water surface in a hot well for storing condensed water and disposed at the lower portion of the heat exchanger, the steams and the incondensible gases are separated, and the incondensible gases are processed in an extraction system. Then, steam condition is improved to an over-heat state, and no large-sized shieldings are necessary. (N.H.)

Temperature monitoring and leak detection in sodium circuits of FBR using Raman distributed fiber optic sensor

International Nuclear Information System (INIS)

Kasinathan, M.; Murali, N.; Sosamma, S.; Babu Rao, C.; Kumar, Anish; Purnachandra Rao, B.; Jayakumar, T.

2013-01-01

This paper discusses the fiber optic temperature sensor based leak detection in the coolant circuits of fast breeder reactor. These sensors measure the temperature based on spontaneous Raman scattering principle and is not influenced by the electromagnetic interference. Various experiments were conducted to evaluate the performance of the fiber optic sensor based leak detection using Raman distributed Temperature Sensor (RDTS). This paper also deals with the details of fiber optic sensor type leak detector layout for the coolant circuit of FBR, performance requirement of leak detection system, description of the test facility, experimental procedure and test results of various experiments conducted. (author)

Assessment of the heat carrier movement in the primary coolant circuit by its own momentum

International Nuclear Information System (INIS)

Kadalev, Stoyan

2014-01-01

Highlights: • We model the heat carrier flow alteration after the circulation pump(s) stop. • The general mathematical model used is described in details. • The model is adapted and applied to a particular example research reactor. • Assessment is presented in detail, step by step with references. • The information provided is enough to apply calculations to another facility. - Abstract: In the presented paper is considered the approach to an assessment of the heat carrier flow alteration in the primary water–water reactor coolant circuit after the circulation pump(s) stop. This topic is highly relevant trough advanced and increased nuclear safety requirements because such a process is observed in case of black-out accident or damaged pump(s). The general mathematical model used is described; enabling preparation of this evaluation adapted and applied to a particular example facility namely a pool type research reactor. The factors influencing to the heat carrier movement by its own momentum are examined. The evaluation measures and includes the factors influencing the heat carrier flow rate from the moment the pump(s) stops down to a negligible value. Assessment is presented in detail, step by step and where needed with references to specific data and/or formulae from reference books to allow repetition of the calculations and/or apply to another facility. The calculations are presented utilizing all necessary data according to the design and technological documentation. No account is given to the pressure of the natural circulation caused by the residual heat generation in the fuel after the reactor scram system extinction of the fission reaction

Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

International Nuclear Information System (INIS)

Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

1996-01-01

The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

Correlation between Ni base alloys surface conditioning and cation release mitigation in primary coolant

Energy Technology Data Exchange (ETDEWEB)

Clauzel, M.; Guillodo, M.; Foucault, M. [AREVA NP SAS, Technical Centre, Le Creusot (France); Engler, N.; Chahma, F.; Brun, C. [AREVA NP SAS, Chemistry and Radiochemistry Group, Paris La Defense (France)

2010-07-01

The mastering of the reactor coolant system radioactive contamination is a real stake of performance for operating plants and new builds. The reduction of activated corrosion products deposited on RCS surfaces allows minimizing the global dose integrated by workers which supports the ALARA approach. Moreover, the contamination mastering limits the volumic activities in the primary coolant and thus optimizes the reactor shutdown duration and environment releases. The main contamination sources on PWR are due to Co-60 and Co-58 nuclides which come respectively Co-59 and Ni-58, naturally present in alloys used in the RCS. Co is naturally present as an impurity in alloys or as the main component of hardfacing materials (Stellites™). Ni is released mainly by SG tubes which represent the most important surface of the RCS. PWR steam generators (SG), due to the huge wetted surface are the main source of corrosion products release in the primary coolant circuit. As corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup, it is of primary importance to gain a better understanding of phenomenon leading to corrosion product release from SG tubes before setting up mitigation measures. Previous studies have shown that SG tubing made of the same material had different release rates. To find the origin of these discrepancies, investigations have been performed on tubes at the as-received state and after exposure to a nominal primary chemistry in titanium recirculating loop. These investigations highlighted the existence of a correlation between the inner surface metallurgical properties and the release of corrosion products in primary coolant. Oxide films formed in nominal primary chemistry are always protective, their morphology and their composition depending strongly on the geometrical, metallurgical and physico-chemical state of the surface on which they

Control circuit for transformer relay

International Nuclear Information System (INIS)

Wyatt, G.A.

1984-01-01

A control circuit for a transformer relay which will automatically momentarily control the transformer relay to a selected state upon energization of the control circuit. The control circuit has an energy storage element and a current director coupled in series and adapted to be coupled with the secondary winding of the transformer relay. A device for discharge is coupled across the energy storage element. The energy storage element and current director will momentarily allow a unidirectional flow of current in the secondary winding of the transformer relay upon application of energy to the control circuit. When energy is not applied to the control circuit the device for discharge will allow the energy storage element to discharge and be available for another operation of the control circuit

Principal working group 3 on primary circuit integrity

Energy Technology Data Exchange (ETDEWEB)

NONE

1993-12-31

The main themes of this conference (13 papers) are: operating experience on leakages and failures in nuclear power plant piping, coolant circuits and steam generator tubes, probabilistic estimation and risk assessment, system failure analysis, leakage events and frequency, leak rate models and crack propagation mechanics, damage mechanisms and rupture probability.

Principal working group 3 on primary circuit integrity

International Nuclear Information System (INIS)

1992-01-01

The main themes of this conference (13 papers) are: operating experience on leakages and failures in nuclear power plant piping, coolant circuits and steam generator tubes, probabilistic estimation and risk assessment, system failure analysis, leakage events and frequency, leak rate models and crack propagation mechanics, damage mechanisms and rupture probability

The chemistry of the X-7 (organic) loop coolant part I, May 1960 to April 1965

International Nuclear Information System (INIS)

Smee, J.L.

1966-01-01

The report describes in detail the X-7 coolant chemistry from the start of loop operation in May 1960 to April 1965. During this period the coolant was Santowax OM containing a nominal 30% high boilers or high molecular weight decomposition products. During the first few months of operation it became apparent that there wa.s a serious problem in the fouling of fuel element heat transfer surfaces. This was overcome by continuous purification of the coolant by Attapulgus clay and filters. Since clay purification has been in use, the fouling rate has been less than 0.2 μg.cm -2 .h -1 (10 μm per year), the target value for successful operation of an organic cooled power reactor. Control of the fouling promoter chlorine has been accomplished by completely excluding it from the vicinity of the loop. Any which does get into the coolant is removed by a bed of Mg ribbon and Pd pellets. Since such a bed has been in use, the Cl content of the coolant has been less than 3 ppm. Also given in this report are: (a) a brief history of the loop since its inception in 1959. (b) the effect of the clay column on the coolant chemistry. (c) a complete description of the current purification, degas and make-up circuits, (d) a summary of the coolant chemistry during all fuel irradiations. (author)

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.

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

Flat plate film cooling at the coolant supply into triangular and cylindrical craters

Directory of Open Access Journals (Sweden)

Khalatov Artem A.

2017-01-01

Full Text Available The results are given of the film cooling numerical simulation of three different schemes including single-array of the traditional round inclined holes, as well as inclined holes arranged in the cylindrical or triangular dimples (craters. The results of simulation showed that at the medium and high values of the blowing ratio (m > 1.0 the scheme with coolant supply into triangular craters improves the adiabatic film cooling efficiency by 1.5…2.7 times compared to the traditional array of inclined holes, or by 1.3…1.8 times compared to the scheme with coolant supply into cylindrical craters. The greater film cooling efficiency with the coolant supply into triangular craters is explained by decrease in the intensity of secondary vortex structures (“kidney” vortex. This is due to the partial destruction and transformation of the coolant jets structure interacting with front wall of the crater. Simultaneously, the film cooling uniformity is increased in the span-wise direction.

Simulation of the dynamic behaviour of the secondary circuit of a WWER-440 type nuclear power plant Pt. 2

International Nuclear Information System (INIS)

Doorenbos, J.; Gacs, A.; Kiss, Zs.

1987-12-01

This report describes the dynamic simulation models of the most important controllers of the secondary circuit of a WWER-440 type nuclear power plant, i.e., the hydraulic turbine controller and the level controls of the condenser hotwell and that of the feedwater tank. Simulation results are also presented. (For dynamic simulation models of the primary circuit of WWER-440 type reactors see Reports KFKI--1983-127 and KFKI--1985-08.) (author) 15 figs

CFD analyses of coolant channel flowfields

Science.gov (United States)

Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

1993-01-01

The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

Dynamic performance investigation of once-through-type steam generator for NPP using a large-scale model

International Nuclear Information System (INIS)

Kats, F.M.; Ostrovskij, L.A.; Ehskin, N.B.

1985-01-01

An experimental bench is described as well as the results of dynamic performance investigation of mass- and heat transfer of the once-through type steam generator for the NPP with weak superheat. Coolant for the primary and secondary circuit is water. Under investigation conditions the possibility of changin.o primary and secondary circuit temperatures has been supphed as well as the primary circuit flow rate and the secondary circuit pressure changes. Transients for differen.t operating conditions are considered. The possibility for construction of the steam generator automatic control system is based

Liquid metal cooled nuclear power plant with direct heat transfer from the primary coolant to the working medium

International Nuclear Information System (INIS)

Hahn, G.

1974-01-01

The cooling systems of the sodium-cooled reactor are entirely inside a containment. The heat transfer from the primary to the secondary coolant - i.e. water - is done in heat exchangers with three-layer tubes. As there is no component cooling heat exchanger, it is advantageous that the layers that are in touch with the primary coolant form part of the wall of the containment. An emergency cooling system inside the containment is also made of three-layer tubes. The tubes of the primary loops have the shape of loops, helices, and spirals surrounding the reactor tank or a biological shield. Between the tubes and the safety wall there are maintenance areas which are accessible from the outside. The three-layer construction prevents a reaction of leaked-out or evaporated sodium with the secondary coolant. (DG) [de

APPLICATION OF MULTIHOLE PRESSURE PROBE FOR RESEARCH OF COOLANT VELOCITY PROFILE IN NUCLEAR REACTOR FUEL ASSEMBLIES

Directory of Open Access Journals (Sweden)

S. M. Dmitriev

2015-01-01

Full Text Available Development of heat and mass transfer intensifiers is a major engineering task in the design of new and modernization of existing fuel assemblies. These devices create lateral mass flow of coolant. Design of intensifiers affects both the coolant mixing and the hydraulic resistance. The aim of this work is to develop a methodology of measuring coolant local velocity in the fuel assembly models with different mixing grids. To solve the problems was manufactured and calibrated multihole pressure probe. The air flow velocity measuring method with multihole pressure probe was used in the experimental studies on the coolant local hydrodynamics in fuel assemblies with mixing grids. Analysis of the coolant lateral velocity vector fields allowed to study the formation of the secondary vortex flows behind the mixing grids, and to determine the basic laws of coolant flow in experimental models. Quantitative data on the coolant flow velocity distribution obtained with a multihole pressure probe make possible to determine the magnitude of the flow lateral velocities in fuel rod gaps, as well as to determine the distance at which damping occurs during mixing. 

Chemistry of the aqueous medium - Determining factor of corrosion in carbon steel components of secondary circuit

International Nuclear Information System (INIS)

Radulescu, M.; Pirvan, I.; Dinu, A.; Velciu, L.

2003-01-01

The interplay of chemistry of aqueous medium and corrosion processes followed by deposition and/or release of corrosion products determines both formation and growth of superficial films as well as the kinetics of ion release from materials into the aqueous medium. Material corrosion in the secondary circuit of a NPP can be minimized by choosing the materials of the components and by a rigorous inspection of the chemistry of aqueous agent. The chemical inspection helps in minimizing: - the corrosion of the components immersed in feedwater and vapor and of Steam Generator components; - 'dirtying' of the systems particularly of the surfaces implied in heat transfer; - the amount of insoluble chemical species resulting in corrosion process and carried along the circuit; - the corrosion of secondary circuit components during revisions or outages. An important role among the chemical parameters of the fluids circulated in NPP tubing appears to be the pH. In CANDU reactors it must be kept within the range of 8.7 to 9.4 by treating the medium with volatile amines (morpholine and cyclohexylamine). A plot is presented giving the corrosion rate of carbon steels as a function of the pH of the medium. Besides, the oxygen concentration dissolved in the aqueous medium must be maintained under 5 μg per water kg. Other factors determining the corrosion rates are also discussed. The paper gives the results of the experiments done with various materials, solutions and analysis methods

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)

Gas turbine with two circuits and intermediate fuel conversion process

International Nuclear Information System (INIS)

Bachl, H.

1978-01-01

The combination of a fuel conversion process with a thermal process saves coolant and subsequent separation plant, in order to achieve the greatest possible use of the mechanical or electrical energy. The waste heat of a thermal circuit is taken to an endothermal chemical fuel conversion process arranged before a second circuit. The heat remaining after removal of the heat required for the chemical process is taken to a second thermal circuit. The reaction products of the chemical process which condense out during expansion in the second thermal process are selectively separated from the remaining gas mixture in the individual turbine stages. (HGOE) [de

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)

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)

Application of liquid chromatography techniques to the measurement of soluble transition metals in PWR primary coolant

International Nuclear Information System (INIS)

Amey, M.D.H.; Brown, G.R.

1987-01-01

Two chromatographic techniques have been developed, and evaluated for the on-line analysis of soluble transition metals, particularly cobalt, in PWR primary coolant. Automatic operation and control, together with data processing and storage has been achieved by interfacing a Dionex ion chromatograph to a microprocessor control system. An absolute detection limit of 0.1 ng cobalt has been obtained which, with on-line sample preconcentration (100 ml), has enabled measurements to be made down to part-per-trillion levels (0.001 ppb). Application of the techniques to PWR coolant analysis was demonstrated by a programme of work on the Half Megawatt Loop at Winfrith. During this work some aspects of the behaviour of soluble metal species have been studied in both de-oxygenated and hydrogenated conditions. The effects of changes in coolant chemistry, operating temperature, and sample line flowrates on circulating impurity levels are reported, together with the dramatic effects observed when part of the circuit pipework was replaced with new stainless steel tubing. (author)

Dimethylamine as a Replacement for Ammonia Dosing in the Secondary Circuit of an Advanced Gas-Cooled Reactor (AGR) Power Station

International Nuclear Information System (INIS)

Armstrong, C.; Mitchell, M.; Bull, A.; Quirk, G.P.; Rudge, A.

2012-09-01

Increasing flow resistance observed over recent years within the helical once-through boilers in the four Advanced Gas-Cooled Reactors (AGRs) at Hartlepool and Heysham 1 Power stations have reduced boiler performance, resulting in reductions in feedwater flow, steam temperatures, power output and the need to carry out periodic chemical cleaning. The root cause is believed to be the development of magnetite deposits with high flow impedance in the 9%Cr evaporator section of the boiler tubing. To prevent continued increases in boiler flow resistance, dimethylamine is being trialled, in one of the four affected units, as a replacement to the conventional ammonia dosing. Dimethylamine increases the pH at temperature around the secondary circuit and, based on full scale boiler rig simulations, is expected to reduce iron transport and prevent flow resistance increases within the evaporator section of the boiler. The dimethylamine plant trial commenced in January 2011 and is ongoing. The feedwater concentration of dimethylamine has been increased progressively towards a final target value of 900 μg kg -1 and its effect on iron transport and boiler pressure loss is being closely monitored. The high steam temperatures (>500 deg. C) of the secondary circuit lead to some decomposition of dimethylamine, which is being carefully monitored at various locations around the circuit. The decomposition products identified with dimethylamine dosing include ammonia, methylamine, formic acid, carbon dioxide and, as yet, unidentified neutral organic species. The effect of dimethylamine dosing on iron transport, boiler pressure drops and its decomposition behaviour around the secondary circuit during the plant trial will be presented in this paper. (authors)

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)

Electrochemical study of galvanic corrosion inhibitors in water-based and ethylene glycol-based heat transfer circuits

International Nuclear Information System (INIS)

Netter, Pierre

1981-01-01

This research thesis reports the search for and the efficiency assessment of mixes of inhibitors for coolant circuits of motor cars. After a discussion of the general properties of water-alcohol solvents (chemical properties, acid-base equilibriums) and of parameters affecting corrosion in coolant circuits, the author proposes an overview of the main inhibitors which are used to protect these circuits against corrosion, and discusses their action mechanism and efficiency. The different methods used to study the corrosion of these circuits are described, and the advantages and drawbacks of test methods are commented. The second part proposes a synthesis of the different corrosion electromechanical mechanisms which may occur with respect to the used metallic materials and to possible galvanic couplings. The next part describes the experimental installations. The last part focuses on the different protections obtained with the different used inhibitor class in terms of results obtained by gravimetric tests and visual examination of samples, current-voltage curves in hydrodynamic regime, and galvanic corrosion tests performed in laboratory or in situ in motor cars [fr

Microcontroller based instrumentation for heater control circuit of tin oxide based hydrogen sensor

International Nuclear Information System (INIS)

Premalatha, S.; Krithika, P.; Gunasekaran, G.; Ramakrishnan, R.; Ramanarayanan, R.R.; Prabhu, E.; Jayaraman, V.; Parthasarathy, R.

2015-01-01

A thin film sensor based on tin oxide developed in IGCAR is used to monitor very low levels of hydrogen (concentration ranging from 2 ppm to 80 ppm). The heater and the sensor patterns are integrated on a miniature alumina substrate and necessary electrical leads are taken out. For proper functioning of the sensor, the heater has to be maintained at a constant temperature of 350°C. The sensor output (voltage signal) varies with H 2 concentration. In fast breeder reactors, liquid sodium is used as coolant. The sensor is used to detect water/steam leak in secondary sodium circuit. During the start up of the reactor, steam leak into sodium circuit generates hydrogen gas as a product that doesn't dissolve in sodium, but escapes to the surge tank containing argon i.e. in cover gas plenum of sodium circuit. On-line monitoring of hydrogen in cover gas is done to detect an event of water/steam leakage. The focus of this project is on the instrumentation pertaining to the temperature control for the sensor heater. The tin oxide based hydrogen sensor is embedded in a substrate which consists of a platinum heater, essentially a resistor. There is no provision of embedding a temperature sensor on the heater surface due to the physical constraints, without which maintaining a constant heater temperature is a complex task

Phenomena occuring in the reactor coolant system during severe core damage accidents

International Nuclear Information System (INIS)

Malinauskas, A.P.

1990-01-01

The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. This paper discusses, how in the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. The physical and chemical processes occurring within the RCS during normal operation of the reactor are relatively uncomplicated and are reasonably well understood. When the flow of coolant is properly adjusted, the thermal energy resulting from nuclear fission (or, in the shutdown mode, from radioactive decay processes) and secondary inputs, such as pumps, are exactly balanced by thermal losses through the RCS boundaries and to the various heat sinks that are employed to effect the conversion of heat to electrical energy. Because all of the heat and mass fluxes remain sensibly constant with time, mathematical descriptions of the thermophysical processes are relatively straightforward, even for boiling water reactor (BWR) systems. Although the coolant in a BWR does undergo phase changes, the phase boundaries remain well-defined and time-invariant

Analogy for Drude's Free Electron Model to Promote Students' Understanding of Electric Circuits in Lower Secondary School

Science.gov (United States)

de Almeida, Maria José B. M.; Salvador, Andreia; Costa, Maria Margarida R. R.

2014-01-01

Aiming at a deep understanding of some basic concepts of electric circuits in lower secondary schools, this work introduces an analogy between the behavior of children playing in a school yard with a central lake, subject to different conditions, rules, and stimuli, and Drude's free electron model of metals. Using this analogy from the first…

Reactor having coolant recycling pump

International Nuclear Information System (INIS)

Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

1991-01-01

In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

Energy Technology Data Exchange (ETDEWEB)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety

2017-06-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

International Nuclear Information System (INIS)

Jobst, Matthias; Wilhelm, Polina; Kliem, Soeren; Kozmenkov, Yaroslav

2017-01-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

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

Formation and hydraulic effects of deposits in high temperature sodium coolant systems

International Nuclear Information System (INIS)

Yunker, W.

1976-01-01

Deposition of sodium impurities in the high temperature (600 0 C), high flow (Reynolds Number approximately equal to 8 x 10 4 ) regions of a sodium coolant circuit is being studied to determine its possible hydraulic effects. Increases in flow impedance (pressure drop/volume flow 2 ) of up to 30 percent have been detected in an annular flow sensor. The apparatus and preliminary results of these tests are presented. Continuing tests are to specifically identify the materials involved and the system conditions under which the formations occur

Process and kinetics of the fundamental radiation-electrochemical reactions in the primary coolant loop of nuclear reactors

International Nuclear Information System (INIS)

Kozomara-Maic, S.

1987-06-01

In spite of the rather broad title of this report, its major part is devoted to the corrosion problems at the RA reactor, i.e. causes and consequences of the reactor shutdown in 1979 and 1982. Some problems of reactor chemistry are pointed out because they are significant for future reactor operation. The final conclusion of this report is that corrosion processes in the primary coolant circuit of the nuclear reactor are specific and that radiation effects cannot be excluded when processes and reaction kinetics are investigated. Knowledge about the kinetics of all the chemical reactions occurring in the primary coolant loop are of crucial significance for safe and economical reactor operation [sr

Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

Energy Technology Data Exchange (ETDEWEB)

Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N. [Moscow Power Engineering Institute (Technical University), Moscow (Russian Federation)

2007-07-01

The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes.

Development of additional module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Bogomazov, D.N.; Poliakov, N.

2007-01-01

The new special module to neutron-physic and thermal-hydraulic computer codes for coolant acoustical characteristics calculation is worked out. The Russian computer code Rainbow has been selected for joint use with a developed module. This code system provides the possibility of EFOCP (Eigen Frequencies of Oscillations of the Coolant Pressure) calculations in any coolant acoustical elements of primary circuits of NPP. EFOCP values have been calculated for transient and for stationary operating. The calculated results for nominal operating were compared with results of measured EFOCP. For example, this comparison was provided for the system: 'pressurizer + surge line' of a WWER-1000 reactor. The calculated result 0.58 Hz practically coincides with the result of measurement (0.6 Hz). The EFOCP variations in transients are also shown. The presented results are intended to be useful for NPP vibration-acoustical certification. There are no serious difficulties for using this module with other computer codes

The influence of the water chemistry regime of the third circuit on the corrosion hydrogen burden to the secondary sodium circuit in the steam generator model of BN-800 reactor

International Nuclear Information System (INIS)

Smykov, V.B.; Ermolaev, N.P.; Kolesnik, A.I.; Egorov, V.A.; Shevchenko, N.N.

1994-01-01

An experimental program was conducted to determine the influence of water chemistry on the corrosion hydrogen burden from the III circuit to the secondary sodium in sodium-heated rig of OTSG of NPP BN-800. Combined water chemistry has given the best passivative effect on steam-generating surfaces and smallest hydrogen burden to secondary sodium during start-up. Common hydrogen increasing in secondary sodium was less then 0.2 ppm. In case of AVT water chemistry (NH 3 +N 2 H 4 ) in III side of OTSG-rig the hydrogen level in secondary sodium was 1.0-1.2 ppm. It means that during first start-up at NPP BN-800 the common hydrogen level in secondary sodium may reaches 0.80-0.85 ppm. 4 figs.; 4 tabs

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

Nuclear reactor cooling device

International Nuclear Information System (INIS)

Hoshi, Masaya; Makihara, Yoshiaki.

1985-01-01

Purpose: To improve the heat transfer performance, as well as reducing and simplifying the structure while preventing the intrusion of primary coolants to utilization systems. Constitution: Heat transfer from the primary coolant circuit to the utilization circuits is conducted by means of heat pipe type heat exchangers. The heat exchanger comprises a tightly closed vessel divided by a partition wall, through which a plurality of heat pipes are passed. The primary coolants receiving the heat from the nuclear reactor enter the first chamber of the heat exchanger to heat the evaporating portion of the heat pipes. The heated flow of steams in the heat pipes transfer to the condensating portion in the second chamber to conduct heat exchange with the utilization system. In this way, since secondary coolant circuits are saved, the heat transfer performance can be improved significantly and the risk of failure can be reduced. (Kamimura, M,)

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.)

Analogy for Drude’s free electron model to promote students’ understanding of electric circuits in lower secondary school

Directory of Open Access Journals (Sweden)

Maria José BM de Almeida

2014-09-01

Full Text Available Aiming at a deep understanding of some basic concepts of electric circuits in lower secondary schools, this work introduces an analogy between the behavior of children playing in a school yard with a central lake, subject to different conditions, rules, and stimuli, and Drude’s free electron model of metals. Using this analogy from the first school contacts with electric phenomena, one can promote students’ understanding of concepts such as electric current, the role of generators, potential difference effects, energy transfer, open and closed circuits, resistances, and their combinations in series and parallel. One believes that through this analogy well-known previous misconceptions of young students about electric circuit behaviors can be overcome. Furthermore, students’ understanding will enable them to predict, and justify with self-constructed arguments, the behavior of different elementary circuits. The students’ predictions can be verified—as a challenge of self-produced understanding schemes—using laboratory experiments. At a preliminary stage, our previsions were confirmed through a pilot study with three classrooms of 9th level Portuguese students.

Ignition circuit for combustion engines

Energy Technology Data Exchange (ETDEWEB)

Becker, H W

1977-05-26

The invention refers to the ignition circuit for combustion engines, which are battery fed. The circuit contains a transistor and an oscillator to produce an output voltage on the secondary winding of an output transformer to supply an ignition current. The plant is controlled by an interrupter. The purpose of the invention is to form such a circuit that improved sparks for ignition are produced, on the one hand, and that on the other hand, the plant can continue to function after loss of the oscillator. The problem is solved by the battery and the secondary winding of the output transformers of the oscillator are connected via a rectifier circuit to produce a resultant total voltage with the ignition coil from the battery voltage and the rectified pulsating oscillator output.

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)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

Energy Technology Data Exchange (ETDEWEB)

Elina Syrjaelahti; Anitta Haemaelaeinen [VTT Processes, P.O.Box 1604, FIN-02044 VTT (Finland)

2005-07-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Hextran-Smabre calculation of the VVER-1000 coolant transient benchmark

International Nuclear Information System (INIS)

Elina Syrjaelahti; Anitta Haemaelaeinen

2005-01-01

Full text of publication follows: The VVER-1000 Coolant Transient benchmark is intended for validation of couplings of the thermal hydraulic codes and three dimensional neutron kinetic core models. It concerns a switching on a main coolant pump when the other three main coolant pumps are in operation. Problem is based on experiment performed in Kozloduy NPP in Bulgaria. In addition to the real plant transient, two extreme scenarios concerning control rod ejection after switching on a main coolant pump were calculated. In VTT the three-dimensional advanced nodal code HEXTRAN is used for the core kinetics and dynamics, and thermohydraulic system code SMABRE as a thermal hydraulic model for the primary and secondary loop. Parallelly coupled HEXTRAN-SMABRE code has been in production use since early 90's, and it has been extensively used for analysis of VVER NPPs. The SMABRE input model is based on the standard VVER-1000 input used in VTT. Last plant specific modifications to the input model have been made in EU projects. The whole core calculation is performed in the core with HEXTRAN. Also the core model is based on earlier VVER-1000 models. Nuclear data for the calculation was specified in the benchmark. The paper outlines the input models used for both codes. Calculated results are introduced both for the coupled core system with inlet and outlet boundary conditions and for the whole plant model. Sensitivity studies have been performed for selected parameters. (authors)

Components of the LWR primary circuit. Pt. 2. Design, construction and calculation. Draft

International Nuclear Information System (INIS)

1995-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 deg C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

Analytical Study on Thermal and Mechanical Design of Printed Circuit Heat Exchanger

Energy Technology Data Exchange (ETDEWEB)

Yoon, Su-Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kim, Eung-Soo [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-09-01

The analytical methodologies for the thermal design, mechanical design and cost estimation of printed circuit heat exchanger are presented in this study. In this study, three flow arrangements of parallel flow, countercurrent flow and crossflow are taken into account. For each flow arrangement, the analytical solution of temperature profile of heat exchanger is introduced. The size and cost of printed circuit heat exchangers for advanced small modular reactors, which employ various coolants such as sodium, molten salts, helium, and water, are also presented.

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.)

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.)

ANALYSIS OF THE IMPACT PROPERTIES OF THE COOLANT RECOVERY SYSTEM HEAT LOSSES OF COMBINED COMPRESSOR-POWER PLANT ON ITS CHARACTERISTICS

Directory of Open Access Journals (Sweden)

Yusha V.L.

2012-12-01

Full Text Available The paper presents results of theoretical analysis of the effectiveness of an ideal thermodynamic cycle internal combustion engine combined with an external utilization of exhaust heat. The influence of the properties of the coolant circuit of utilization on its operational parameters and characteristics of the power plant.

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

Gas turbine electric generator

International Nuclear Information System (INIS)

Nemoto, Masaaki; Yuhara, Tetsuo.

1993-01-01

When troubles are caused to a boundary of a gas turbine electric generator, there is a danger that water as an operation medium for secondary circuits leaks to primary circuits, to stop a plant and the plant itself can not resume. Then in the present invention, helium gases are used as the operation medium not only for the primary circuits but also for the secondary circuits, to provide so-called a direct cycle gas turbine system. Further, the operation media of the primary and secondary circuits are recycled by a compressor driven by a primary circuit gas turbine, and the turbine/compressor is supported by helium gas bearings. Then, problems of leakage of oil and water from the bearings or the secondary circuits can be solved, further, the cooling device in the secondary circuit is constituted as a triple-walled tube structure by way of helium gas, to prevent direct leakage of coolants into the reactor core even if cracks are formed to pipes. (N.H.)

Thermal hydraulic analysis of aggressive secondary cooldown in a small break loss of coolant accident with a total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, Seok Jung; Lim, Ho Gon; Yang, Joon Eon

2003-03-01

Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). The present study focuses on detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model using RIA for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study in this year is to evaluate the success cri-teria of Aggressive Secondary Cooldown (ASC) in a Small Size Loss Of Coolant Accident (SBLOCA) without HPSI and to enhance the understanding of related thermal hydraulic behavior and phenomena. An effort was made to evaluate the system success criteria and a mission time for the recovery action by an operator to prevent the core damage for that accident scenario. The accident scenario for KSNP was a 2 inch coldleg break LOCA with a total loss of High Pressure Safety Injection (HPSI) and 1/2 Low Pressure Safety Injection (LPSI) available and perform-ing ASC limited by 55.6 .deg. C/hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip. It successively reached the LPSI condition for about 1.5hr after starting the ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria of 1204.4 .deg. C (2200 .deg. F). Sensitivity studies were performed for (1) cool-ant average temperature parameters, (2) ASC operation control method, (3) operation start time, (4) 1 inch break size. The present analysis identified thermal hydraulic phenomena and parameters affecting on the behavior, which consist of coolant break flow and inventory, parameters governing secondary heat removal, ASC operation control method, and its reference temperature parameters. In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that an operator should maintain the ade-quate ASC operation. However, it is necessary to evaluate the uncertainties arisen from the

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

New equivalent lumped electrical circuit for piezoelectric transformers.

Science.gov (United States)

Gonnard, Paul; Schmitt, P M; Brissaud, Michel

2006-04-01

A new equivalent circuit is proposed for a contour-vibration-mode piezoelectric transformer (PT). It is shown that the usual lumped equivalent circuit derived from the conventional Mason approach is not accurate. The proposed circuit, built on experimental measurements, makes an explicit difference between the elastic energies stored respectively on the primary and secondary parts. The experimental and theoretical resonance frequencies with the secondary in open or short circuit are in good agreement as well as the output "voltage-current" characteristic and the optimum efficiency working point. This circuit can be extended to various PT configurations and appears to be a useful tool for modeling electronic devices that integrate piezoelectric transformers.

Deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Parsons, P.D.; Hindle, E.D.; Mann, C.A.

1986-09-01

The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail.

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.)

Limit the effects of secondary circuit water or steam piping breaks in the reactor building

International Nuclear Information System (INIS)

Nachev, N.

2001-01-01

The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

Effects of Coolant Temperature Changes on Reactivity for Various Coolants in a Liquid Salt Cooled Very High Temperature Reactor (LS-VHTR)

International Nuclear Information System (INIS)

Casino, William A. Jr.

2006-01-01

The purpose of this study is to perform an investigation into the relative merit of various salts and salt compounds being considered for use as coolants in the liquid salt cooled very high temperature reactor platform (LS-VHTR). Most of the non-nuclear properties necessary to evaluate these salts are known, but the neutronic characteristics important to reactor core design are still in need of a more extensive examination. This report provides a two-fold approach to further this investigation. First, a list of qualifying salts is assembled based upon acceptable non-nuclear properties. Second, the effect on system reactivity for a secondary system transient or an off-normal or accident condition is examined for each of these salt choices. The specific incident to be investigated is an increase in primary coolant temperature beyond normal operating parameters. In order to perform the relative merit comparison of each candidate salt, the System Temperature Coefficient of Reactivity is calculated for each candidate salt at various state points throughout the core burn history. (author)

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

General thermodynamic description of pollutants and preservatives in water at high temperature: application to primary and secondary circuits of power plants

International Nuclear Information System (INIS)

Alvarez, Jorge L.; Kukuljan, Juan A.; Gutkowski, Karin; Japas, Maria L.; Fernandez Prini, Roberto

1999-01-01

A formalism has been developed for the description of solubilities and other thermodynamic functions, based only on the Krichevskii function and properties of the pure solvent. This formalism is applied to pollutants of primary and secondary circuits nuclear power plants. (author)

Study of the accumulation and distribution of the radioactivity in the cooling circuit of the BOR-60 reactor

International Nuclear Information System (INIS)

Kizin, V.D.; Konyashov, V.V.; Lisitsyn, E.S.; Polyakov, V.I.; Chechetkin, Yu.V.

1976-04-01

The results of measurements of the radioactivity of the coolant and the deposits in the primary circuit of the BOR-60 reactor during its five years of operation are discussed. The values calculated for the exposure dose rate from the piping system and the contribution of the γ-radiation from the corrosion and fission product nuclides are given. The efficiency of coolant draining from the pipes in reducing the dose rate is estimated. (orig.) [de

Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

International Nuclear Information System (INIS)

Deniskin, V.; Nalivaev, V.; Fedik, I.; Vishnevski, U.; Dmitriev, A.

2003-01-01

Full text: The work that has been conducted so far justifies a possibility of constructing a reactor with a non-traditional coolant to develop radically new reactors and their cycles with perfect architecture. A solid coolant, for example, the carbon-based one, allows to design the primary circuit of nuclear reactor without excess pressure. Such coolant withstands temperatures up to ∼4000 deg. K without a collapse. The analysis of theory and experiments produced requirements to be met by a solid coolant used in the primary circuit of nuclear reactor. One of the most important requirements is the arrangements for a continuous and homogeneous gravity flow of the coolant through all core sections taking into account the dust caused by wear and some amount of fractured particles. Therefore, the idea is that the mass of particles should resemble a liquid to a certain extend. The particles should be sphere like with average diameter from 0.5 to 2.0 mm and nonsphericity rate not more than 10%. 'Angle of repose' of particles to the horizon can be utilised as a validity criterion of particles which should not exceed 25 deg. The heat transfer coefficient should be increased up to the practical maximum value. In 1996 - 1997 the system of experimental facilities were built in the Scientific and Research Institute 'Luch' to prove the possibility to reliably cool a nuclear reactor with a flow of solid particles and to obtain a minimum set of data for the conceptual design of such reactor with solid coolant. The facility allows the research of the flow stability, heat mass transfer in the core, lifetime wearing of particles of the solid coolant. In 1994-1999 5 batches of particles of different size were fabricated in accordance to different technologies. Four batches were graphite-based and one was aluminium oxide-based (Al 2 O 3 ). The purpose was to verify how the heat transfer coefficient was changing as the particle size varied. The average diameter of graphite particles

Computer modelling the potential benefits of amines in NPP Bohunice secondary circuit

International Nuclear Information System (INIS)

Fountain, M.J.; Smiesko, I.

1998-01-01

The use of computer modelling of PWR and WWER secondary circuit chemistry was already demonstrated in the past. The model was used to illustrate the technical and economic advantages, compared with ammonia, of using an 'advanced', high basicity, low volatility amines to raise the liquid phase pH(T) in the moisture separator and other areas swept by wet steam. Since the 1995, this technique has been successfully applied to a number of power plants and the computer model has been progressively developed. This paper describes the preliminary results of an ongoing assessment being carried out for the VVER 440 plants at Bohunice. The work for Bohunice is being funded by the 'Know How Fund', a department in the British Government's Foreign and Commonwealth Office. (J.P.N.)

A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

Energy Technology Data Exchange (ETDEWEB)

Lister, D. [University of New Brunswick, Fredericton, NB (Canada). Dept. of Chemical Engineering; Lang, L.C. [Atomic Energy of Canada Ltd., Chalk River Lab., ON (Canada)

2002-07-01

Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

A mechanistic model for predicting flow-assisted and general corrosion of carbon steel in reactor primary coolants

International Nuclear Information System (INIS)

Lister, D.

2002-01-01

Flow-assisted corrosion (FAC) of carbon steel in high-temperature lithiated water can be described with a model that invokes dissolution of the protective oxide film and erosion of oxide particles that are loosened as a result. General corrosion under coolant conditions where oxide is not dissolved is described as well. In the model, the electrochemistry of magnetite dissolution and precipitation and the effect of particle size on solubility move the dependence on film thickness of the diffusion processes (and therefore the corrosion rate) away from reciprocal. Particle erosion under dissolving conditions is treated stochastically and depends upon the fluid shear stress at the surface. The corrosion rate dependence on coolant flow under FAC conditions then becomes somewhat less than that arising purely from fluid shear (proportional to the velocity squared). Under non-dissolving conditions, particle erosion occurs infrequently and general corrosion is almost unaffected by flow For application to a CANDU primary circuit and its feeders, the model was bench-marked against the outlet feeder S08 removed from the Point Lepreau reactor, which furnished one value of film thickness and one of corrosion rate for a computed average coolant velocity. Several constants and parameters in the model had to be assumed or were optimised, since values for them were not available. These uncertainties are no doubt responsible for the rather high values of potential that evolved as steps in the computation. The model predicts film thickness development and corrosion rate for the whole range of coolant velocities in outlet feeders very well. In particular, the detailed modelling of FAC in the complex geometry of one outlet feeder (F11) is in good agreement with measurements. When the particle erosion computations are inserted in the balance equations for the circuit, realistic values of crud level are obtained. The model also predicts low corrosion rates and thick oxide films for inlet

Correlating activity incorporation with properties of oxide films formed on material samples exposed to BWR and PWR coolants in Finnish nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Bojinov, M.; Kinnunen, P.; Laitinen, T.; Maekelae, K.; Saario, T.; Sirkiae, P. [VTT Industrial Systems, Espoo (Finland); Buddas, T.; Halin, M.; Kvarnstroem, R.; Tompuri, K. [Fortum Power and Heat Oy, Loviisa Power Plant, Loviisa (Finland); Helin, M.; Muttilainen, E.; Reinvall, A. [Teollisuuden Voima Oy, Olkiluoto (Finland)

2002-07-01

The extent of activity incorporation on primary circuit surfaces in nuclear power plants is connected to the chemical composition of the coolant, to the corrosion behaviour of the material surfaces and to the structure and properties of oxide films formed on circuit surfaces due to corrosion. Possible changes in operational conditions may induce changes in the structure of the oxide films and thus in the rate of activity incorporation. To predict these changes, experimental correlations between water chemistry, oxide films and activity incorporation, as well as mechanistic understanding of the related phenomena need to be established. In order to do this, flow-through cells with material samples and facilities for high-temperature water chemistry monitoring have been installed at Olkiluoto unit 1 (BWR) and Loviisa unit 1 (PWR) in spring 2000. The cells are being used for two major purposes: To observe the changes in the structure and activity levels of oxide films formed on material samples exposed to the primary coolant. Correlating these observations with the abundant chemical and radiochemical data on coolant composition, dose rates etc. collected routinely by the plant, as well as with high-temperature water chemistry monitoring data such as the corrosion potentials of relevant material samples, the redox potential and the high-temperature conductivity of the primary coolant. We describe in this paper the scope of the work, give examples of the observations made and summarize the results on oxide films that have been obtained during one full fuel cycle at both plants. (authors)

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

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)

Components of the LWR primary circuit. Pt. 2. Komponenten des Primaerkreises von Leichtwasserreaktoren. T. 2

Energy Technology Data Exchange (ETDEWEB)

1984-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400/sup 0/C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

The deformation, oxidation and embrittlement of PWB fuel cladding in a loss-of-coolant accident

International Nuclear Information System (INIS)

Parsons, P.D.; Hindle, E.D.; Mann, C.A.

1986-09-01

The scope of this report is limited to the oxidation, embrittlement and deformation of PWB fuel in a loss of coolant accident in which the emergency core coolant systems operate in accordance with the design, ie accidents within the design basis of the plant. A brief description is given of the thermal hydraulic events during large and small breaks of the primary circuit, followed by the correct functioning and remedial action of the emergency core cooling systems. The possible damage to the fuel cladding during these events is also described. The basic process of oxidation of zircaloy-4 fuel cladding by steam, and the reaction kinetics of the oxidation are reviewed in detail. Variables having a possible influence on the oxidation kinetics are also considered. The embrittlement of zircaloy-4 cladding by oxidation is also reviewed in detail. It is related to fracture during the thermal shock of rewetting or by the ambient impact forces as a result of post-accident fuel handling. Criteria based both on total oxidation and on the detailed distribution of oxygen through the oxidised cladding wall are considered. The published computer codes for the calculation of oxygen concentration are reviewed in terms of the model employed and the limitations apparent in these models when calculating oxygen distribution in cladding in the actual conditions of a loss of coolant accident. The factors controlling the deformation and rupture of cladding in a loss of coolant accident are reviewed in detail. (author)

Thermal hydraulic analyses of accidents associated with coolant leak from the primary circuit through a hole 10 mm equivalent diameter for the needs of PTS

International Nuclear Information System (INIS)

Krhounkova, J.; Kral, P.; Parduba, Z.

1999-10-01

The conservative assumptions of the analyses were oriented towards a worsening of the process with respect to the pressurized thermal shock (PTS). Four variants were treated, viz. leaks from the cold or hot leg, each at the rated power or zero power. Since the temperature of water supplied to the primary circuit by the emergency core cooling system is an important parameter with respect to a PTS, the calculations were performed by the iterative procedure: the basic thermal hydraulic calculation was performed by the RELAP5/MOD3.2.1 code which calculates the behaviour of the primary and secondary circuits, whereas the MELCOR code was used to calculate the behaviour of the parameters in the hermetic rooms. The calculation by the RELAP code was then repeated using data from the MELCOR calculations. Interventions by the reactor operators were also considered. (P.A.)

Secondary circuit water chemistry and related problems with SG

Energy Technology Data Exchange (ETDEWEB)

Ignatov, V; Ivanov, V [Balakovo Nuclear Power Plant (Russian Federation)

2001-07-01

Necessity for SG feed water and blowdown systems modernization Balakovo NPP steam generators PGV-1000M was identified at Units with VVER-1000 during commissioning separational, thermo-hydraulic and thermo-chemical testings. It was discovered, that in zone of 'hot' header coolant salt concentration (concentration of dissolved salts) was almost 2 times more, than salt concentration in blowdown water. A number of chemical testings was performed to investigate and optimize salts distribution in water volume of PGV-1000. (R.P.)

Secondary circuit water chemistry and related problems with SG

International Nuclear Information System (INIS)

Ignatov, V.; Ivanov, V.

2001-01-01

Necessity for SG feed water and blowdown systems modernization Balakovo NPP steam generators PGV-1000M was identified at Units with VVER-1000 during commissioning separational, thermo-hydraulic and thermo-chemical testings. It was discovered, that in zone of 'hot' header coolant salt concentration (concentration of dissolved salts) was almost 2 times more, than salt concentration in blowdown water. A number of chemical testings was performed to investigate and optimize salts distribution in water volume of PGV-1000. (R.P.)

Sustainability issues in circuit board recycling

DEFF Research Database (Denmark)

Legarth, Jens Brøbech; Alting, Leo; Baldo, Gian Luca

1995-01-01

The resource recovery and environmental impact issues of printed circuit board recycling by secondary copper smelters are discussed. Guidelines concerning material selection for circuit board manufacture and concerning the recycling processes are given to enhance recovery efficiency and to lower...

Secondary coolant circuit for liquid-metal cooled reactor and steam generator for such a circuit

International Nuclear Information System (INIS)

Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.; Traiteur, R.; Zuber, T.

1984-01-01

An upper buffer tank and downstream buffer tank are disposed inside the steam generators. The downstream briffer tank is annular and it surrounds and communicates with a zone of the steam generator through which the liquid metal flows towards the bottom between the exchange zone and the outlet nozzle. The pressure of the inert gas blanket in the downstream buffer volume is more important than this one in the upper buffer volume. The invention applies to fast neutron nuclear reactor cooled by sodium [fr

Primary circuit and reactor core T-H characteristics determination of WWER 440 reactors

International Nuclear Information System (INIS)

Hermansky, J.; Petenyi, V.; Zavodsky, M.

2010-01-01

The WWER-440 nuclear fuel vendor permanently improves the assortment of produced nuclear fuel assemblies for achieving better fuel cycle economy and reactor operation safety. During unit refuelling there also could be made some other changes in hydraulic parameters of primary circuit (change of impeller wheels, hydraulic resistance coefficient changes of internal parts of primary circuit, etc.). Therefore it is necessary to determine real coolant flow rate through the reactor during units start-up after their refuelling, and also to have the skilled methodology and computing code for analyzing factors, which affecting the inaccuracy of coolant flow redistribution determination through reactor on flows through separate parts of reactor core in any case of parallel operation of different assembly types. Computing code TH-VCR and CORFLO are used for reactor core characteristics determination for one type of fuel and control assemblies and also in case of parallel operation of different assembly types. The code TH-VCR is able to calculate coolant flow rate for different combinations of three different fuel assembly channel types and three different control assembly channel types. The CORFLO code deals the area of the reactor core which consists of 312 fuel assemblies and 37 control assemblies. Regarding the rotational 60 deg symmetry of reactor core only 1/6 of reactor core with 59 fuel assemblies is taken into account. Computing code CORFLO is verified and validated at this time. Paper presents some results from measurements of coolant flow rate through reactors during start-up after unit refuelling and short description of computing code TH-VCR and CORFLO with some calculated results. (Authors)

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)

Fission product retention during faults involving steam generator tube rupture

International Nuclear Information System (INIS)

Rodliffe, R.S.

1983-08-01

In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

Distribution and behavior of tritium in the Coolant-Salt Technology Facility

International Nuclear Information System (INIS)

Mays, G.T.; Smith, A.N.; Engel, J.R.

1977-04-01

A 1000-MW(e) Molten-Salt Breeder Reactor (MSBR) is expected to produce 2420 Ci/day of tritium. As much as 60 percent of the tritium produced may be transported to the reactor steam system (assuming no retention by the secondary coolant salt), where it would be released to the environment. Such a release rate would be unacceptable. Experiments were conducted in an engineering-scale facility--the Coolant-Salt Technology Facility (CSTF)--to examine the potential of sodium fluoroborate, the proposed coolant salt for an MSBR, for sequestering tritium. The salt was believed to contain chemical species capable of trapping tritium. A series of 5 experiments--3 transient and 2 steady-state experiments--was conducted from July of 1975 through June of 1976 where tritium was added to the CSTF. The CSTF circulated sodium fluoroborate at temperatures and pressures typical of MSBR operating conditions. Results from the experiments indicated that over 90 percent of tritium added at steady-state conditions was trapped by sodium fluoroborate and appeared in the off-gas system in a chemically combined (water-soluble) form and that a total of approximately 98 percent of the tritium added at steady-state conditions was removed through the off-gas system overall

TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

International Nuclear Information System (INIS)

Chalhoub, E.S.

1980-09-01

A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

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.)

Analysis of an accident with the main circulation tube rupture at the WWER-1000

International Nuclear Information System (INIS)

Boyadzhiev, A.I.; Stefanova, S.J.

1984-01-01

In connection with the forthcoming construction of a npp with the wwer-1000 reactor the loss of coolant accident associated with the main circulation tube rupture at the inlet near the reactor is analyzed. The relap4/mod6 program is used for the analysis. The data obtained show that the coolant outflow stage continues for about 25s. On the average the pressure in the circuits varies from 16 to 10 mpa per 0.1s and then it continues to decrease slowly. The pressure in the steam generator at the secondary circuits end increases approximately up to 6.9 MPa as a result of steam generator blocking and remaining coolant heating and then somewhat decreases owing to the primary circuit cooling. By the end of the fuel and can temperatures are equalized and the heat transfer coefficient is stabilized at the level of 100 w/1 (m 2 xK). It is concluded that during a loss of coolant accident at the wwer-1000 reactor in procesess of coolant blowdown in the medium power fuel elemets neither the fuel, melting temperature (3000 k), nor the critical temperature (1000 k) of plastic deformation zirconiu can initiation are attained

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

Flow boiling test of GDP replacement coolants

International Nuclear Information System (INIS)

Park, S.H.

1995-01-01

The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

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

Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

International Nuclear Information System (INIS)

Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

2015-01-01

In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

30 CFR 75.900-2 - Approved circuit schemes.

Science.gov (United States)

2010-07-01

... device installed in the main secondary circuit at the source transformer may be used to provide undervoltage protection for each circuit that receives power from that transformer. (c) One circuit breaker may... accordance with the settings listed in the tables of the National Electric Code, 1968. ...

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.)

Study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP

International Nuclear Information System (INIS)

Soares, Alexandre de Souza

2014-01-01

The present paper proposes a study of a loss of coolant accident of a PWR reactor through a Full Scope Simulator and computational code RELAP. To this end, it considered a loss of coolant accident with 160 cm 2 breaking area in cold leg of 20 circuit of the reactor cooling system of nuclear power plant Angra 2, with the reactor operating in stationary condition, to 100% power. It considered that occurred at the same time the loss of External Power Supply and the availability of emergency cooling system was not full. The results obtained are quite relevant and with the possibility of being used in the planning of future activities, given that the construction of Angra 3 is underway and resembles the Angra 2. (author)

Coolant cleanup method in a nuclear reactor

International Nuclear Information System (INIS)

Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

1983-01-01

Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

Method of operation for a nuclear power plant

International Nuclear Information System (INIS)

Barleon, L.; Dalle Donne, M.; Dorner, S.

1976-01-01

A method for the operation of a nuclear power plant is described whose primary coolant circuit contains liquid metal as coolant, and for whose secondary circuit which is coupled to the first by means of heat exchangers, one or more chemical compositions dissociating at higher temperatures and recombining at lower temperatures or associating into the original types are used as operating medium. N 2 O 4 is suggested as operating medium, and metal halogenides e.g. Al 2 Cl 6 , Al 2 Br 6 are also cited as suitable. Corrosion dangers are regarded as low. (UWI) [de

Noise diagnostics research at the Institute for Electric Power Research

International Nuclear Information System (INIS)

Voeroess, L.

1983-01-01

The main goal of recent research is the designing of a diagnostic system for the Paks nuclear power plant. It includes the designing of measuring cascades for measuring mechanical vibrations caused by flow and pressure fluctuations in the primary coolant circuit, the arrangement of measuring cascades for measuring turbine bearing vibrations in the secondary coolant circuit, and the development, manufacturing and commissioning of a central data acquisition and processing unit for the whole diagnostic system. Simultaneously computer models are developed for improving the interpretation. Moreover, it is planned to develop methods for estimating residual lifetime and the status of plant components. (author)

Multirods burst tests under loss-of-coolant conditions

International Nuclear Information System (INIS)

Kawasaki, S.; Uetsuka, H.; Furuta, T.

1983-01-01

In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Sikora, D

1984-02-28

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

Extended Life Coolant Testing

Science.gov (United States)

2016-06-06

number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

Loss of Coolant Accidents (LOCA): Study of CAREM Reactor Response

International Nuclear Information System (INIS)

Gonzalez, Jose; Gimenez, Marcelo

2000-01-01

We analyzed the neutronic and thermohydraulic response of CAREM25 reactor and the safety systems involved in a Loss Of Coolant Accident (LOCA).This parametric analysis considers several break diameters (1/2inch, 3/4inch, 1inch, 1.1/2inch and 2inches) in the vapor zone of the Reactor Pressure Vessel.For each accidental sequence, the successful operation of the following safety systems is modeled: Second Safety System (SSS), Residual Heat Removal System (RHRS) and Safety Injection System (SIS). Availability of only one module is postulated for each system.On the other hand, the unsuccessful operation of all safety systems is postulated for each accidental sequence.In both cases the First Shutdown System (FSS) actuates, and the loss of Steam Generator secondary flow and Chemical and Control of Volume System (CCVS) unavailability are postulated.Maximum loss of coolant flow, reactor power and time for safety systems operation are analyzed, as well as its set point parameters.We verified that safety systems are dimensioned to satisfy the 48 hours cooling criteria

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

Water quality control program in experimental circuits

International Nuclear Information System (INIS)

Cegalla, Miriam A.

1996-01-01

The Water Quality Control Program of the Experimental Circuits visualizes studying the water chemistry of the cooling in the primary and secondary circuits, monitoring the corrosion of the systems and studying the mechanism of the corrosion products transport in the systems. (author)

Importance of ECP in the prediction of radiation fields in PWR and VVER primary circuits

International Nuclear Information System (INIS)

Urquidi-Macdonald, M.; Jacesko, S.L.; Macdonald, Digby D.; Salter-Williams, M.

2002-01-01

A model has been developed for predicting mass and activity transport in the primary coolant circuits of PWRs and VVERs with the objective of demonstrating and quantifying the importance of the electrochemical corrosion potential (ECP) in determining the impact of both processes on reactor operation. The model initially employs a radiolysis/mixed potential code to calculate the ECP at four locations (core, hot leg, steam generator, cold leg) and the ECP is then used to estimate the local magnetite solubility. The solubility is then averaged around the loop to yield the ''background'' solubility. Comparison of the background solubility with the local solubility determines whether precipitation or dissolution will occur at any given point in the circuit under any given set of conditions. It is further assumed that the concentration of 59 Co in the coolant is given by the isotopic fraction of this species compared with iron averaged over all materials and weighted by the respective wetted areas. Activation of 59 Co to 60 Co is assumed to occur in the coolant phase by fast, epithermal, and thermal neutron capture. The calculated activity is then used to train an artificial neural network (ANN) to establish relationships between activity at any given location and the operating properties of the reactor, including coolant pH, ECP, temperature, power level, etc. The model predicts that during shut down, magnetite (and hence 59 Co) migrates to the core, where it is irradiated and activated, particularly during subsequent start-up. During start-up, the magnetite (and hence 60 Co) migrates from the core to out-of-core surfaces where it establishes the radiation fields. (authors)

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

Sizewell 'B' power station public inquiry: CEGB proof of evidence

International Nuclear Information System (INIS)

Dutton, L.M.

1982-11-01

A description is given of the main sources of radioactivity and the barriers which would prevent their release at the proposed Sizewell B nuclear power station. The methods of calculating the radiological consequences associated with the most significant faults within the design are outlined. The following possibilities are considered: faults involving high fuel temperatures; loss of coolant accidents; secondary coolant circuit faults; waste processing system faults; irradiated fuel faults. (U.K.)

Emergency cooling of presurized water reactor

International Nuclear Information System (INIS)

Sykora, D.

1981-01-01

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

The compact simulator for Tihange nuclear plant

International Nuclear Information System (INIS)

Gueben, M.

1982-01-01

After an introduction about the simulators for nuclear plants, a description is given of the compact simulator for the Tihange nuclear power plant as well as the simulated circuits and equipments such as the primary and secondary coolant circuits. The extent of simulation, the functions used by the instructor, the use of the simulator, the formation programme and construction planning are described. (AF)

Passive surveillance: a technique to characterize the condition of power and control circuits in a nuclear plant

International Nuclear Information System (INIS)

Meininger, R.D.; Dinsel, M.R.

1985-01-01

This paper reports on progress by EG and G Idaho in examination of electrical circuits exposed to the accident environment at Three Mile Island Unit 2 (TMI-2) during and after the loss-of-coolant accident of March 28, 1979. Interpretations of the data collected suggest that the major functional impact on the electrical circuits (a) occurs very late in time, (b) is caused by moisture intrusion, and (c) can be detected by remote surveillance prior to functional failure. The electrical testing was performed from outside the TMI-2 Reactor Building at the penetrations using a special circuit characterization and diagnostic system developed by EG and G Idaho. This paper concentrates on representative data from those circuits which were recently retested. 12 refs., 9 figs

Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

Science.gov (United States)

Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

2017-10-01

The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

Analysis on transient hydrodynamic characteristics of cavitation process for reactor coolant pump

International Nuclear Information System (INIS)

Wang Xiuli; Wang Peng; Yuan Shouqi; Zhu Rongsheng; Fu Qiang

2014-01-01

The reactor coolant pump hydrodynamic characteristics at different cavitation conditions were studied by using flow field analysis software ANSYS CFX, and the corresponding data were processed and analyzed by using Morlet wavelet transform and fast Fourier transform. The results show that gas content presents the law of exponential function with the pressure reduction or time increase. In the cavitation primary condition, the pulsation frequency of head for the reactor coolant pump is mainly low frequency, and the main frequency of pressure pulsation is still rotation frequency while the effect of the pressure pulsation caused by cavitation on main frequency is not obvious. With the development of cavitation, the pressure fluctuation induced by cavitation becomes more serious especially for the main frequency, secondary frequency and pulsating amplitude while the head pulsation frequency is given priority to low frequency pulse. Under serious cavitation condition, the head pulsation frequency is given priority to irregular changes of pulse high frequency, and also contains almost regular changes of low frequency. (authors)

Upper internals of PWR with coolant flow separator

International Nuclear Information System (INIS)

Chevereau, G.; Heuze, A.

1989-01-01

The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

Evaluation of an optimized coolant circuit conception in a thermal whole vehicle environment with respect to the consumption of primary energy; Bewertung eines optimierten Kuehlmittelkreislaufkonzeptes in einer thermischen Gesamtfahrzeugumgebung hinsichtlich des Primaerenergieverbrauchs

Energy Technology Data Exchange (ETDEWEB)

Schulze, Mirko; Neumann, Alexander; Tilch, Benjamin; Eilts, Peter [Technische Univ. Braunschweig (Germany). Inst. fuer Verbrennungskraftmaschinen; Niedersaechsisches Forschungszentrum Fahrzeugtechnik (NFF), Braunschweig (Germany)

2012-11-01

This work deals with a co-simulation vehicle environment developed by the institute of internal combustion engines (ivb) of the Technical University Braunschweig as a tool to analyze the thermal effects in the power train during the warm-up phase, especially on the fuel consumption. This allows evaluating new drive train concepts in early stages of development by using power train thermal management techniques (TMM). Therefore you are able to give an objective statement for these techniques by analyzing the changes in fuel consumption. The used simulation models will be introduced and the mechanical and thermal behavior is verified using test bench data. An optimized coolant circuit concept in GT Suite {sup registered}, developed at the institute is identified and coupled to a thermal engine model. In this paper, the potentials for reducing primary energy consumption in the New European Driving Cycle (NEDC) are presented. (orig.)

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.)

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

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)

Flow accelerated corrosion and life management of the secondary circuit of the Embalse nuclear power plant

International Nuclear Information System (INIS)

Chocron, Mauricio; La Gamma, Ana M.; Fernandez, Narciso; Moyano, Ricardo; Schiersmann, Christian; Ovando, Luis E.; Sainz, Ricardo A.; Keitelman, Alberto

2003-01-01

Flow accelerated corrosion is a matter of concern in secondary circuits of nuclear power plants as well as in fossil fired plants. It contributes to the piping wall thinning and to the corrosion products transport to the steam generators. Because it is a generalized corrosion phenomena, could address to extensive failures. In that sense the plants conduct extensive programs of surveillance of piping degradation. Because the problem involves many variables like alloys, water chemistry and hydrodynamics several models have been proposed in the literature. In the present paper the variables have been organized in a spreadsheet which allows the calculation of normalized risk factors. (author)

Modification of the motor control system of secondary cooling pumps of the multipurpose reactor - GA Siwabessy (RSG-GAS) to overcome the electrical flicking

International Nuclear Information System (INIS)

Yayan Andriyanto; Tufiq, M.; Adin Sudirman; Koes Indrakoesoema

2012-01-01

Modification of the motor control system of secondary cooling pumps that have been carried out to overcome the electrical flicking. Consequences of unanticipated flicking is the cooling pumps will shut down. Modification is done by replacing the control system PLN to Uninterruptible Power Supply (UPS), plus a component of Miniature Circuit Breaker (MCB) 220 Volt 2 Amp 2 Pole, 2 Auxiliary Contactor-type 3RH1131-1APOO with 1 NO auxiliary contact (normally Open) and two time-1BP30-3RP1505 type relay (off delay). Delay time to overcome the electrical flicking is set to 1 seconds and 2 seconds, taking into calculate the electrical flicking that occurred while in the setting of time limits, the predicted secondary cooling water pressure inside the pipeline is still eligible and the motor cooling system operation can still continue to operate. In the flicking of an electrical incident on 13 April 2010 at 0:45 Pm showed that the modifications of the secondary coolant pump motors operation when the reactor operates with a power of 15 MW. (author)

Development of lead-bismuth coolant technology for nuclear device

International Nuclear Information System (INIS)

Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

2004-01-01

Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

Device for preventing coolant outflow in a reactor

International Nuclear Information System (INIS)

Nemoto, Kiyomitsu; Mochizuki, Keiichi.

1975-01-01

Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

Purification of inert gas circuits of nuclear power facilities from tritium and hydrogen

International Nuclear Information System (INIS)

Eichler, R.

1985-08-01

Removing hydrogen and tritium from the inert primary coolant of a high temperature reactor is very important in regard to the process heat disposition. In this work a gas purification for a high temperature module reactor was laid out constructionally and researched technically. This system removes the contamination of the primary circuit with the aid of chemical getter beds of Cer alloy particles. (orig./PW) [de

Method of charging instruments into liquid metal coolant

International Nuclear Information System (INIS)

Yamazaki, Hiroshi

1980-01-01

Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

14C Behaviour in PWR coolant

International Nuclear Information System (INIS)

Sims, Howard; Dickinson Shirley; Garbett, Keith

2012-09-01

Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

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.)

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.)

Components of the primary circuit of LWRs. Design, construction and calculation. Komponenten des Primaerkreises von Leichtwasserreaktoren. Auslegung, Konstruktion und Berechnung

Energy Technology Data Exchange (ETDEWEB)

1980-01-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

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.)

The monitoring of oxygen, hydrogen and carbon in the sodium circuits of the PFR

International Nuclear Information System (INIS)

Mason, L.; Morrison, N.S.; Robertson, C.M.; Trevillion, E.A.

1984-01-01

The paper reviews the instrumentation available for monitoring oxygen, hydrogen, tritium and carbon impurity levels on the primary and secondary circuits of PFR. Circuit oxygen levels measured using electrochemical oxygen meters are compared to estimates from circuit plugging meters. The data are interpreted in the light of information from cold trap temperatures. Measurements of secondary circuit hydrogen levels using both the sodium and gas phase hydrogen detection equipment are compared to estimates of circuit hydrogen levels from plugging meters and variations in sodium phase hydrogen levels during power operation are discussed. (author)

A Preliminary Study of Banana Stem Juice as a Plant-Based Coagulant for Treatment of Spent Coolant Wastewater

Directory of Open Access Journals (Sweden)

Habsah Alwi

2013-01-01

Full Text Available The effectiveness of banana stem juice as a natural coagulant for treatment of spent coolant wastewater was investigated . Three main parameters were studied, namely, chemical oxygen demand (COD, suspended solids (SSs, and turbidity of effluent. Coagulation experiments using jar test were performed with a flocculation system where the effects of spent coolant wastewater pH as well as banana stem juice dosage on coagulation effectiveness were examined. The highest recorded COD, SS, and turbidity removal percentages by banana stem juice were 80.1%, 88.6%, and 98.5%, respectively, observed for effluent at pH 7 using 90 mL dosage. The inulin concentration in the banana stem was examined to be 1.22016 mg/mL. It could be concluded that banana stem juice showed tremendous potential as a natural coagulant for water treatment purposes and could be applied in the pretreatment stage of Malaysian spent coolant wastewater prior to secondary treatment.

Impact of load follow operation on the chemistry of the primary and secondary circuit of a pressurized water reactor

International Nuclear Information System (INIS)

Boettcher, F.; Riehm, S.; Bolz, M.; Speck, A.

2012-09-01

products activity concentrations in the loop surfaces and the dose rates at the loop lines that are monitored and assessed regularly. The secondary circuit is operated according the H-AVT chemistry (high-pH, treatment with volatile chemicals) to minimize the release of metals from carbon- and low alloy steels. The impact of load follow operation on the corrosion protection of the carbon- and low alloy steels on the secondary side due to changes of physico-chemical conditions is discussed. (authors)

The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

International Nuclear Information System (INIS)

Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

1993-01-01

Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

Development of fast-burn combustion with elevated coolant temperatures for natural gas engines. Final report, May 1985-May 1990

Energy Technology Data Exchange (ETDEWEB)

Bruch, K.L.; Dennis, J.W.

1990-09-01

The overall objective of the work was to improve the state of the art in the gas fired spark ignited engine for use in a cogeneration system. Four characteristics were enhanced for cogeneration, namely, Low Pressure Gas Induction, Improved Shaft Thermal Efficiency, Low NOx Emissions, and Increased Jacket Coolant Temperature. Using Taguchi methods and statistical design of experiment methodologies, an engine design evolved that exhibited: The ability to run satisfactorily on supply gas pressure as low as 1.5 psig (goal: 1 psig); A brake specific fuel consumption as low as 6950 Btu/hp-hr (36.6% thermal efficiency) at 2 gm/hp-hr NOx (goal: 7000 acceptable, 6800 excellent with NOx no more than 2 gm/hp-hr); A jacket water coolant system (with oil cooler on the same circuit) temperature of 225 F (goal); and The ability to burn gas with Methane Number as low as 67 (goal).

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

Decontamination of main coolant pumps

International Nuclear Information System (INIS)

Roofthooft, R.

1988-01-01

Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

Major activated corrosion products cobalt, silver and antimony in the primary coolant of PWR power plants

International Nuclear Information System (INIS)

Xu Mingxia

2012-01-01

The production of the major activated corrosion products such as cobalt, silver and antimony in the primary coolant of PWR power plants and the impacts on the increase of the dose rates caused by these corrosion products during the shutdown are described in the paper. Investigating the corrosion product behavior during the operation and shutdown periods aims at detecting the appearance of these radiological pollutants in the early time and searching relevant solutions that may enable eventually to decrease the dose rate. The solutions may include: Replacing critical material in the primary system's equipment and components, which contact with primary coolant circuit to possibly limit the source term, Elaborating strictly the specific chemical and shutdown procedure to optimize the purification capacity and to minimize the over-contaminations; Improving purification techniques according to the real operation circumstance, and limiting the impacts of these pollutants. It is obvious in the real practices that implementing appropriate solution will be benefit to decrease or limit the pollutants species like cobalt, silver and antimony. (author)

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

Predicting the conditions under which vibroacoustic resonances with external periodic loads occur in the primary coolant circuits of VVER-based NPPs

Science.gov (United States)

Proskuryakov, K. N.; Fedorov, A. I.; Zaporozhets, M. V.

2015-08-01

The accident at the Japanese Fukushima Daiichi nuclear power plant (NPP) caused by an earthquake showed the need of taking further efforts aimed at improving the design and engineering solutions for ensuring seismic resistance of NPPs with due regard to mutual influence of the dynamic processes occurring in the NPP building structures and process systems. Resonance interaction between the vibrations of NPP equipment and coolant pressure pulsations leads to an abnormal growth of dynamic stresses in structural materials, accelerated exhaustion of equipment service life, and increased number of sudden equipment failures. The article presents the results from a combined calculation-theoretical and experimental substantiation of mutual amplification of two kinds of external periodic loads caused by rotation of the reactor coolant pump (RCP) rotor and an earthquake. The data of vibration measurements at an NPP are presented, which confirm the predicted multiple amplification of vibrations in the steam generator and RCP at a certain combination of coolant thermal-hydraulic parameters. It is shown that the vibration frequencies of the main equipment may fall in the frequency band corresponding to the maximal values in the envelope response spectra constructed on the basis of floor accelerograms. The article presents the results from prediction of conditions under which vibroacoustic resonances with external periodic loads take place, which confirm the occurrence of additional earthquake-induced multiple growth of pressure pulsation intensity in the steam generator at the 8.3 Hz frequency and additional multiple growth of vibrations of the RCP and the steam generator cold header at the 16.6 Hz frequency. It is shown that at the elastic wave frequency equal to 8.3 Hz in the coolant, resonance occurs with the frequency of forced vibrations caused by the rotation of the RCP rotor. A conclusion is drawn about the possibility of exceeding the design level of equipment vibrations

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

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

Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

Effects of addition glycerol co-product of biodiesel in the thermophysical properties of water-glycerol solution applied as secondary coolant

Energy Technology Data Exchange (ETDEWEB)

Medeiros, Pedro Samuel Gomes; Barbosa, Cleiton Rubens Formiga; Fontes, Francisco de Assis Oliveira [Federal University of Rio Grande do Norte, Natal, RN (Brazil). Energy Laboratory. Thermal Systems Studies Group], e-mail: cleiton@ufrnet.br

2010-07-01

This paper evaluates the effects of glycerol concentration on thermophysical properties of water-glycerol solution applied as a secondary coolant in refrigeration systems by expansion-indirect. The processing of triglycerides for biodiesel production generates glycerol as co-product and there are concerns of environmental and economic order on the surplus of glycerol. The addition of glycerol in water alters the colligative and thermophysical properties (melting point, mass, specific heat, thermal conductivity and dynamic viscosity). There are studies that prove the feasibility of using glycerol as an additive and this paper has the goal to verify the changes on properties compared with pure water. This comparison was made from data obtained by the software simulation and they analyzed using graphs and tables. It was shown that glycerol increases the density and dynamic viscosity, and reduces the specific heat and thermal conductivity. This behavior of water-glycerol solution is proportional to the mass concentration of glycerol and it is justified because the glycerol has low values of specific heat, thermal conductivity and high viscosity when compared with water. Despite the losses in the thermophysical properties, glycerol shows its potential application, because of the cryoscopic effect and it is a non-toxic substance at low cost. (author)

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

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.

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

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.)

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

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Jones, R.D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

Science.gov (United States)

Jones, Robert D.

1978-01-01

A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

Reactor coolant system hydrostatic test and risk analysis for the first AP1000 unit

International Nuclear Information System (INIS)

Cao Hongjun; Yan Xiuping

2013-01-01

The cold hydrostatic test scheme of the primary coolant circuit, of the first AP1000 unit was described. Based on the up-stream design documents, standard specifications and design technical requirements, the select principle of test boundary was identified. The design requirements for water quality, pressure, temperature and temporary hydro-test pump were proposed. A reasonable argument for heating and pressurization rate, and cooling and depressurization rate was proposed. The possible problems and risks during the hydrostatic test were analyzed. This test scheme can provide guidance for the revisions and implementations of the follow-up test procedures. It is a good reference for hydrostatic tests of AP1000 units in the future in China. (authors)

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.)

Real-time reactor coolant system pressure/temperature limit system

International Nuclear Information System (INIS)

Newton, D.G.; Schemmel, R.R.; Van Scooter, W.E. Jr.

1991-01-01

This patent describes an system, used in controlling the operating of a nuclear reactor coolant system, which automatically calculates and displays allowable reactor coolant system pressure/temperature limits within the nuclear reactor coolant system based upon real-time inputs. It comprises: means for producing signals representative of real-time operating parameters of the nuclear reactor cooling system; means for developing pressure and temperature limits relating the real-time operating parameters of the nuclear reactor coolant system, for normal and emergency operation thereof; means for processing the signals representative of real-time operating parameters of the nuclear reactor coolant system to perform calculations of a best estimate of signals, check manual inputs against permissible valves and test data acquisition hardware for validity and over/under range; and means for comparing the representative signals with limits for the real-time operating parameters to produce a signal for a real-time display of the pressure and temperature limits and of the real-time operating parameters use an operator in controlling the operation of the nuclear reactor coolant system

Comparison between MAAP and ECART predictions of radionuclide transport throughout a French standard PWR reactor coolant system; Transport des radionucleides dans le circuit primaire d`un REP: comparaison des codes MAAP et ECART

Energy Technology Data Exchange (ETDEWEB)

Hervouet, C.; Ranval, W. [Electricite de France (EDF), 92 - Clamart (France); Parozzi, F.; Eusebi, M. [Ente Nazionale per l`Energia Elettrica, Rome (Italy)

1996-04-01

In the framework of a collaboration agreement between EDF and ENEL, the MAAP (Modular Accident Analysis Program) and ECART (ENEL Code for Analysis of radionuclide Transport) predictions about the fission product retention inside the reactor cooling system of a French PWR 1300 MW during a small Loss of Coolant Accident were compared. The volatile fission products CsI, CsOH, TeO{sub 2} and the structural materials, all of them released early by the core, are more retained in MAAP than in ECART. On the other hand, the non-volatile fission products, released later, are more retained in ECART than in MAAP, because MAAP does not take into account diffusion-phoresis: in fact, this deposition phenomenon is very significant when the molten core vaporizes the water of the vessel lower plenum. Centrifugal deposition in bends, that can be modeled only with ECART, slightly increases the whole retention in the circuit if it is accounted for. (authors). 18 refs., figs., tabs.

Loss-of-coolant accident for large pipe breaks in light water reactor plants

International Nuclear Information System (INIS)

Keusenhoff, J.

1980-01-01

The importance of loss-of-coolant accidents (LOCA) and their control for nuclear reactor safety is explained. Showing the cooling circuits and emergency core cooling systems (ECCS) of both, PWR and BWR, the possible break spectrum and the general sequence of events is discussed. The governing physical phenomena for the different LOCA phases are pointed out in more detail. Special emphasis is taken on rules, regulations and failure criteria for licensing purposes. Analysis methods and codes for both, evaluation and best-estimate model are compared under deterministic and probabilistic approach, respectively. Some insight in present integral and separate effect tests demonstrates the interdependency of analysis and experiment. Results of LOCA analysis and experiments show the present state of the art. (orig.)

ECCS control circuit

International Nuclear Information System (INIS)

Sato, Takashi.

1986-01-01

Purpose: To afford a sufficient margin to pressure vibrations upon starting of an automatic depressurization system by dispersing pressure vibration in suppression water due to the opening action of an automatic releaf valve in the automatic depressurization system thereby reducing the dynamic load exerted to the surface of the suppression walls. Constitution: Upon occurrence of loss of coolant accidents, an automatic releaf valve for automatic depressurization is opened to deliver the steams in the pressure vessel into the suppression pool. Since a plurality of automatic releaf valves have usually been disposed, if they are opened simultaneously, excess dynamic loads are exerted due to the pressure vibrations to the wall surface of the suppression pool. In this invention, a control circuit is disposed such that the opening timing for each of the automatic releaf valves is deviated upon occurrence of a driving signal for the automatic depressurization system to thereby disperse the pressure vibrations in the suppression water. (Kamimura, M.)

Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

Energy Technology Data Exchange (ETDEWEB)

Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

2009-06-15

To prevent the appearance of the conditions for resonance interaction between the fluid flow and the reactor internals (RI), fuel rod (FR ) and fuel assemblies (FA) it is necessary to de-tune Eigen frequency of coolant pressure oscillations (EFCPO) and natural frequency of mechanical element's oscillations and also of the system which is formed by the comprising of these elements. Other words it is necessary to de-tune acoustic resonance frequency and natural frequencies of RI, FR and FA. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of the coolant outside of which there is no resonant interaction with structure vibrations. The presented work is devoted to finding the solution of this problem. There are results of an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. Abnormal growth of intensity of pressure pulsations in a mode with definite value of reactor capacity have been found out by measurements on VVER - 1000 reactor. This phenomenon has been found out casually and its original reason had not been identified. Paper shows that disappearance of this effect could be reached by realizing outlet of EFCPO from so-called, pass bands of frequencies (PBF). PBF is located symmetrical on both parties from frequency of own oscillations of FA. Methods, algorithms of calculations and quantitative estimations are developed for EFCPO, Q and PBF in various modes of operation NPP with VVER-1000. Results of calculations allow specifying area of resonant interaction EFCPO with vibrations of FR, FA and a basket of reactor core. For practical realization of the received results it is offered to make corresponding additions to the design documentation and maintenance instructions of the equipment of the NPP with VVER-1000. The improvement of these documents

Primary coolant recycling device for FBR type reactor

International Nuclear Information System (INIS)

Kanbe, Mitsuru; Tokiwai, Moriyasu

1998-01-01

A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

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)

Water chemistry and corrosion in water-steam circuits of nuclear power plants

International Nuclear Information System (INIS)

Gardent, R.; Menet, O.

1981-01-01

The water and steam circuits of steam generators in pressurized-water nuclear power plants are described together with the mechanism of denting, and the corrosion of spacer plates that leads to cracks in tubes by constriction. The different chemical specifications applicable to the water of the secondary circuit of the generators in normal operation and on first commissioning are listed. The results obtained and the measurements of chemical values taken in operation on the water in the secondary circuits of steam generators at Fessenheim and Bugey are presented [fr

A study of the large break loss-of-coolant accident in the Angra-1 nuclear power plant

International Nuclear Information System (INIS)

Borges, E.M.

1984-01-01

The simulation of the Angra-I nuclear power plant under the condition of large break loss of coolant accident is presented, the thermal-hydraulic analysis of the primary circuit during each phase of the acident and thermal analysis of the hottest fuel rod curing reflooding are shown. Computer codes RELAP4/MOD5 (options EM and FLOOD) and TOODEE 2 are used to perform these computations. Fuel rod peak temperatures reached during the simulation are below the permissible levels. However, during the reflooding phase; the maximum oxidation of the cladding exceeds the limit of 0.17 times the original cladding thickness. (Author) [pt

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

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%

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

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

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

Investigation and evaluation of erosion-corrosion status of the secondary side steam-water system at Paks Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Schunk, J.; Patek, G.; Pinter, T.; Baracska Varju, I.; Doma, A.; Kovacs, A.; Nemeth, P.; Tilky, P. [Paks Nuclear Power Plant Ltd (Hungary); Osz, J. [Budapest Univ. of Science and Economy (Hungary)

2009-07-01

There were four identical VVER-440 type units installed at Paks Nuclear Power Plant with 440 MW electrical capacity of each, between 1982 and 1987. Based on inherent reserve capacity of these units, the operational and maintenance experiences and examples of power upgrading abroad it was decided to increase the power of our units in a stepwise way. Presently, this upgrading project is close to finalization resulting in 500 MW electrical capacity of each unit. To realize the planned electrical power increase it was necessary to increase the primary and secondary heat power, which requested the increase of mass flow, temperature and pressure of secondary coolant flowing through the steam generators. According to the preliminary expert evaluations these increased parameters would not cause such an increase in humidity of steam leaving the turbine houses that could adversely impair our long term power upgrading plans. The steam humidity was determined on our units and the values did not even exceed the original design values. In spite of these preliminary investigations, significant accumulation of erosion-corrosion products at different places of secondary circuit was found and erosion damages of some secondary side equipment were discovered as a probable consequence of power upgrading. An extensive evaluation programme has been started involving experts and institutes of material testing, chemistry and hydrodynamics. Results and details of that work are given in our presentation. (authors)

ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2005-01-01

The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

Full reactor coolant system chemical decontamination qualification programs

Energy Technology Data Exchange (ETDEWEB)

Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-03-01

Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

Steam as turbine blade coolant: Experimental data generation

Energy Technology Data Exchange (ETDEWEB)

Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

1995-10-01

Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

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

Development of a steady-state calculation model for the KALIMER PDRC(Passive Decay Heat Removal Circuit)

International Nuclear Information System (INIS)

Chang, Won Pyo; Ha, Kwi Seok; Jeong, Hae Yong; Kwon, Young Min; Eoh, Jae Hyuk; Lee, Yong Bum

2003-06-01

A sodium circuit has usually featured for a Liquid Metal Reactor(LMR) using sodium as coolant to remove the decay heat ultimately under accidental conditions because of its high reliability. Most of the system codes used for a Light Water Reactor(LWR) analysis is capable of calculating natural circulation within such circuit, but the code currently used for the LMR analysis does not feature stand alone capability to simulate the natural circulation flow inside the circuit due to its application limitation. To this end, the present study has been carried out because the natural circulation analysis for such the circuit is realistically raised for the design with a new concept. The steady state modeling is presented in this paper, development of a transient model is also followed to close the study. The incompressibility assumption of sodium which allow the circuit to be modeled with a single flow, makes the model greatly simplified. Models such as a heat exchanger developed in the study can be effectively applied to other system analysis codes which require such component models

Development of computational program for studying the reactor control system in PWR plants; Desenvolvimento de um programa computacional para estudo do sistema de controle do reator em plantas PWR

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Jose Ricardo de; Soares, Adalberto Jose [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)

2002-07-01

In this work a computational program is presented which has been developed for specific application on the study of the reactor control system of a typical PWR plant. As to the basic function of simulating power transients the program has the following structure: a representative mathematical model of the dynamic and stationary behaviors of the primary circuit; a group of equations associated to the reactor power control and system pressure control; screens for the entry of reference data as well as of control blocks and control bar speed programming module parameters; main entering screens for the configuration of the excitement/transient function as well as of simulation time and control mood; and graphical output of all the process variables incorporated to the model. As premise it has been considered as sufficient the modeling of the primary circuit, a differential equation being used which associates the average temperature of the coolant within the steam generator with the potency transferred to the secondary circuit, denominated 'secondary potency', as an interface with the secondary circuit. Every transient - ramp or step - is established upon the 'turbine power' variable, which in turn is related to the 'secondary power' variable by means of a differential equation that represents a first - order delay, having adjustable parameters on the data - entry screen. In the neutronic model as defined for the reactor, the reactivity feedback effects due to primary circuit pressure variation, as well as fuel and coolant temperature variation, were taken into consideration. Thermo-hydraulics constants and project data taken from the available bibliography, adapted to a particular small PWR unit conception , were employed for loading the program. With the open-loop simulation results a positive qualitative evaluation of the program was obtained, in comparison to published results related to simulators bearing equal purposes, more

Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

Directory of Open Access Journals (Sweden)

Lefèvre Grégory

2012-01-01

Full Text Available In the primary circuit of pressurized water reactors (PWR, the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek theory and used as such to interpret this industrial phenomenon.

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

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

Components of the primary circuit of LWRs. Design, construction and calculation. Draft. Komponenten des Primaerkreises von Leichtwasserreaktoren. Auslegung, Konstruktion und Berechnung. Entwurf

Energy Technology Data Exchange (ETDEWEB)

1979-10-01

This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673/sup 0/K (400/sup 0/C). The primary circuit as the pressure continement of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding off from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

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.)

Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

International Nuclear Information System (INIS)

Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

2009-01-01

In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

Reliability improvement of potential transformer and secondary circuit of 6 kV 1E-class buses in Qinshan nuclear power plant

International Nuclear Information System (INIS)

Qian Houjun

2014-01-01

There are design defects in potential transformer (PT) and secondary circuit of 6 kV 1E-class buses in Qinshan Nuclear Power Plant Nuclear Island. During the operating period, there happened several serious operational events (loss of power) caused by PT resonance. The essay analyses the defects of original design, and put forward corresponding modification measures, which have been carried out by two steps between 2009 and 2010, and after the modification the same problems have not happened again. (author)

3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

Energy Technology Data Exchange (ETDEWEB)

Ramajo, Damian, E-mail: dramajo@santafe-conicet.gov.ar; Corzo, Santiago; Schiliuk, Nicolas; Nigro, Norberto

2013-12-15

A computational fluid dynamics (CFD) simulation of the reactor pressure vessel (RPV) of the pressurized heavy water reactor (PHWR) of 745 electrical MW Atucha II nuclear power plant was carried out. A three dimensional (3D) detailed model was employed to simulate coolant circuit considering the upper and lower plenums, the downcomer and the hot and cold legs. Control rods and coolant channel tubes at the upper plenum were included to quantify the mixing flow with more realism. The whole set of 451 coolant channels were modeled by means of a zero dimensional methodology. That is, the effect of each coolant channel was modeled through the introduction of a source point at the upper plenum and a sink point at the lower plenum. For each coupled sink/source points (SSP) the mass, momentum and energy balance were solved considering the local pressure difference and the temperature between the corresponding points where sinks and sources were placed. Based on this strategy, three models with increasingly level of approximation were implemented. For the first model the 451 coolant channels were reduced to only 57 pairs of SSP to represent all the coolant channels, concentrating the effect of several coolant channels in a unique pair of sink and source while taking into account geometric design details. For the second model, 225 pairs of SSP were introduced. Finally, for the third model each one of the 451 coolant channels were modeled by means of one pair of SSP. Depending on the coolant channel location, the radial power distribution and the pressure loss caused by the corresponding flow restrictor present by design were considered. Simulations carried out give insight in the complexity of the flow. As expected, the greater the details of the model the better the accuracy reached in the representation of the RPV behavior. In addition, the flow distributor located at the lower plenum showed to be very efficient since, the mass flow at each channel was found to be fairly

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

Mathematical model of the reactor coolant pump

International Nuclear Information System (INIS)

Kozuh, M.

1989-01-01

The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

Inspection of the Sizewll 'B' reactor coolant pump flywheels

International Nuclear Information System (INIS)

McNulty, A.L.; Cheshire, A.

1992-01-01

The Sizewell ''B'' safety case has categorised some primary circuit items as components for which failure is considered to be incredible. These Incredibility of Failure (IOF) components are particularly critical in their safety function, and specially stringent and all embracing provisions are made in their design, manufacture, inspection and operation. These provisions are such as to limit the probability of failure to levels which are so low that it does not have to be taken into account and no steps are necessary to control the consequences. The reactor coolant pump flywheel is considered to be an IOF component. Consequently there is a need for rigorous inspection during both manufacture and in service (ISI). The ISI requirement results in the need for an automated inspection. There is therefore a prerequisite to perform a Pre-Service Inspection (PSI) for baseline fingerprinting purposes. Furthermore there is a requirement that the inspection procedure, the inspection equipment and the operators are validated at the Inspection Validation Centre (IVC) of the AEA Technology laboratories at Risley. Development work is described. (author)

On-Line Coolant Chemistry Analysis

International Nuclear Information System (INIS)

LM Bachman

2006-01-01

Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

Monitoring Sodium Circuits and ACSR cables using Fiber Optic Sensors

International Nuclear Information System (INIS)

Kasinathan, M.; Sosamma, S.; Babu-Rao, C.; Kumar, Anish; Purna-Chandra-Rao, B.; Murali, N; Jayakumar, T.

2013-06-01

Raman Distributed Temperature Sensors (RDTS) are attractive for the monitoring of coolant loop systems in nuclear power plants and monitoring of overhead power transmission lines. This paper discusses deployment of RDTS on double walled pipelines of primary sodium circuits in Fast Breeder Reactors (FBR). It is demonstrated as a proof-of-concept on a test loop with water as the leaking medium. Path delay multiplexing is adopted to improve the spatial resolution from 1.02 m to 0.5 m. A second application focuses on the influence of environmental factors on the detectability of defects in the ACSR cables using RDTS. (authors)

A programming language for composable DNA circuits.

Science.gov (United States)

Phillips, Andrew; Cardelli, Luca

2009-08-06

Recently, a range of information-processing circuits have been implemented in DNA by using strand displacement as their main computational mechanism. Examples include digital logic circuits and catalytic signal amplification circuits that function as efficient molecular detectors. As new paradigms for DNA computation emerge, the development of corresponding languages and tools for these paradigms will help to facilitate the design of DNA circuits and their automatic compilation to nucleotide sequences. We present a programming language for designing and simulating DNA circuits in which strand displacement is the main computational mechanism. The language includes basic elements of sequence domains, toeholds and branch migration, and assumes that strands do not possess any secondary structure. The language is used to model and simulate a variety of circuits, including an entropy-driven catalytic gate, a simple gate motif for synthesizing large-scale circuits and a scheme for implementing an arbitrary system of chemical reactions. The language is a first step towards the design of modelling and simulation tools for DNA strand displacement, which complements the emergence of novel implementation strategies for DNA computing.

Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

International Nuclear Information System (INIS)

Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

2000-01-01

A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

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

Quantitative secondary electron detection

Science.gov (United States)

Agrawal, Jyoti; Joy, David C.; Nayak, Subuhadarshi

2018-05-08

Quantitative Secondary Electron Detection (QSED) using the array of solid state devices (SSD) based electron-counters enable critical dimension metrology measurements in materials such as semiconductors, nanomaterials, and biological samples (FIG. 3). Methods and devices effect a quantitative detection of secondary electrons with the array of solid state detectors comprising a number of solid state detectors. An array senses the number of secondary electrons with a plurality of solid state detectors, counting the number of secondary electrons with a time to digital converter circuit in counter mode.

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)

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.

Safety aspects of water chemistry in light water reactors

International Nuclear Information System (INIS)

1988-12-01

The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

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

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

Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

International Nuclear Information System (INIS)

Carter, G.W.

1979-01-01

Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

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

BWR fuel assembly bottom nozzle with one-way coolant flow valve

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1987-01-01

In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

33 CFR 183.435 - Conductors in circuits of 50 volts or more.

Science.gov (United States)

2010-07-01

... HOMELAND SECURITY (CONTINUED) BOATING SAFETY BOATS AND ASSOCIATED EQUIPMENT Electrical Systems Manufacturer... more. (c) This section does not apply to communication systems; electronic navigation equipment; resistance conductors that control circuit amperage; conductors in secondary circuits of ignition systems...

Actively controlling coolant-cooled cold plate configuration

Science.gov (United States)

Chainer, Timothy J.; Parida, Pritish R.

2015-07-28

A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

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

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.

Transient two-phase performance of LOFT reactor coolant pumps

International Nuclear Information System (INIS)

Chen, T.H.; Modro, S.M.

1983-01-01

Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

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

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

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

Improvement of Measurement Accuracy of Coolant Flow in a Test Loop

Energy Technology Data Exchange (ETDEWEB)

Hong, Jintae; Kim, Jong-Bum; Joung, Chang-Young; Ahn, Sung-Ho; Heo, Sung-Ho; Jang, Seoyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

In this study, to improve the measurement accuracy of coolant flow in a coolant flow simulator, elimination of external noise are enhanced by adding ground pattern in the control panel and earth around signal cables. In addition, a heating unit is added to strengthen the fluctuation signal by heating the coolant because the source of signals are heat energy. Experimental results using the improved system shows good agreement with the reference flow rate. The measurement error is reduced dramatically compared with the previous measurement accuracy and it will help to analyze the performance of nuclear fuels. For further works, out of pile test will be carried out by fabricating a test rig mockup and inspect the feasibility of the developed system. To verify the performance of a newly developed nuclear fuel, irradiation test needs to be carried out in the research reactor and measure the irradiation behavior such as fuel temperature, fission gas release, neutron dose, coolant temperature, and coolant flow rate. In particular, the heat generation rate of nuclear fuels can be measured indirectly by measuring temperature variation of coolant which passes by the fuel rod and its flow rate. However, it is very difficult to measure the flow rate of coolant at the fuel rod owing to the narrow gap between components of the test rig. In nuclear fields, noise analysis using thermocouples in the test rig has been applied to measure the flow velocity of coolant which circulates through the test loop.

A procedure for temperature-stress fields calculation of WWER-1000 primary circuit in PTS event

Energy Technology Data Exchange (ETDEWEB)

Petkov, G [Technical Univ., Dept. Thermal and Nuclear Power Engineering, Sofia (Bulgaria); Groudev, P; Argirov, J [Bulgarian Academy of Science, Inst. for Nuclear Research and Nuclear Energy, Sofia (Bulgaria)

1997-09-01

The paper presents the procedure of an investigation of WWER-1000 primary circuit temperature-stress field by the use of thermohydraulic computation data for a pressurized thermal shock event ``Core overcooling``. The procedure is based on a model of the plane stress state with ideal contact between wall and medium for the calculation. The computation data are calculated on the base of WWER-1000 thermohydraulic model by the RELAP5/MOD3 codes. This model was developed jointly by the Bulgarian and BNL/USA staff to provide an analytical tool for performing safety analysis. As a result of calculations by codes the computation data for temperature field law (linear laws of a few distinguished parts) and pressure of coolant at points on inner surface of WWER-1000 primary circuit equipment are received. Such calculations can be used as a base for determination of all-important load-carrying sections of the primary circuit pipes and vessels, which need further consideration. (author). 7 refs, 2 figs, 2 tabs.

High-Voltage MOSFET Switching Circuit

Science.gov (United States)

Jensen, Kenneth A.

1995-01-01

Circuit reliably switches power at supply potential of minus 1,500 V, with controlled frequency and duty cycle. Used in argon-plasma ion-bombardment equipment for texturing copper electrodes, as described in "Texturing Copper To Reduce Secondary Emission of Electrons" (LEW-15898), also adapted to use in powering gaseous flash lamps and stroboscopes.

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

International Nuclear Information System (INIS)

Kanzleiter, T.F.

1976-01-01

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

Trace organics in AGR coolants

International Nuclear Information System (INIS)

Smith, R.; Green, L.O.; Johnson, P.A.V.

1980-01-01

Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

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

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.

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.

Acoustical environment of gas-cooled nuclear reactors

International Nuclear Information System (INIS)

Blevins, R.D.

1986-01-01

Methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound, are described. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed. (author)

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.

ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

Directory of Open Access Journals (Sweden)

Veijo KAUPPINEN

2012-11-01

Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

Improving Coolant Effectiveness through Drill Design Optimization in Gundrilling

Science.gov (United States)

Woon, K. S.; Tnay, G. L.; Rahman, M.

2018-05-01

Effective coolant application is essential to prevent thermo-mechanical failures of gun drills. This paper presents a novel study that enhances coolant effectiveness in evacuating chips from the cutting zone using a computational fluid dynamic (CFD) method. Drag coefficients and transport behaviour over a wide range of Reynold numbers were first established through a series of vertical drop tests. With these, a CFD model was then developed and calibrated with a set of horizontal drilling tests. Using this CFD model, critical drill geometries that lead to poor chip evacuation including the nose grind contour, coolant hole configuration and shoulder dub-off angle in commercial gun drills are identified. From this study, a new design that consists a 20° inner edge, 15° outer edge, 0° shoulder dub-off and kidney-shaped coolant channel is proposed and experimentally proven to be more superior than all other commercial designs.

A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

International Nuclear Information System (INIS)

Yoo, J. W.; Kim, S. J.; Kim, Y. I.

2000-01-01

A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

Heat pump augmentation of nuclear process heat

International Nuclear Information System (INIS)

Koutz, S.L.

1986-01-01

A system is described for increasing the temperature of a working fluid heated by a nuclear reactor. The system consists of: a high temperature gas cooled nuclear reactor having a core and a primary cooling loop through which a coolant is circulated so as to undergo an increase in temperature, a closed secondary loop having a working fluid therein, the cooling and secondary loops having cooperative association with an intermediate heat exchanger adapted to effect transfer of heat from the coolant to the working fluid as the working fluid passes through the intermediate heat exchanger, a heat pump connected in the secondary loop and including a turbine and a compressor through which the working fluid passes so that the working fluid undergoes an increase in temperature as it passes through the compressor, a process loop including a process chamber adapted to receive a process fluid therein, the process chamber being connected in circuit with the secondary loop so as to receive the working fluid from the compressor and transfer heat from the working fluid to the process fluid, a heat exchanger for heating the working fluid connected to the process loop for receiving heat therefrom and for transferring heat to the secondary loop prior to the working fluid passing through the compressor, the secondary loop being operative to pass the working fluid from the process chamber to the turbine so as to effect driving relation thereof, a steam generator operatively associated with the secondary loop so as to receive the working fluid from the turbine, and a steam loop having a feedwater supply and connected in circuit with the steam generator so that feedwater passing through the steam loop is heated by the steam generator, the steam loop being connected in circuit with the process chamber and adapted to pass steam to the process chamber with the process fluid

Corrosion product behaviour in the primary circuit of the KNK nuclear reactor facility

International Nuclear Information System (INIS)

Stamm, H.H.; Stade, K.Ch.

1976-01-01

During nuclear operation of the KNK facility from 1972 until September 1974 the composition and behaviour of radionuclides occuring in the primary circuit were investigated. Besides traces of 140 Ba/ 140 La, no fission product activity was detectable in the KNK primary circuit. The fuel element purification from sodium deposits (prior to transport to the reprocessing plant) did not yield any indication of a fuel element failure during KNK-I operation. The activity inventory of the primary loop was exclusively made up of activated corrosion products and 22 Na. The main activity was due to 65 Zn, followed by 54 Mn, 22 Na, sup(110m)Ag, 182 Ta, 60 Co and 124 Sb. It was found that the sorption of 65 Zn and 54 Mn on crucibles made from nickel was condiserably higher than on vessels made from other materials. This observation was confirmed both in tests with material samples from the primary circuit and for disks of gate valves of the primary circuit. sup(110m)Ag did hardly exhibit any sorption effects and had been dissolved largely homogeneously in the hot primary coolant. In the first primary cold trap which was removed from the circuit after some 20,000 hours of operation, only 65 Zn and 54 Mn were detected in addition to traces of 60 Co and 182 Ta. (author)

Design technology development of the main coolant pump for an integral reactor

International Nuclear Information System (INIS)

Park, J. S.; Lee, J. S.; Kim, M. H.; Kim, D. W.; Kim, J. I.

2004-01-01

All of the reactor coolant pump currently used in commercial nuclear power plant were imported from foreign country. Now, the developing program of design technology for the reactor coolant pump will be started in a few future by domestic researchers. At this stage, the design technology of the main coolant pump for an integral reactor is developed based on the regulation of domestic nuclear power plant facilities. The main coolant pump is a canned motor axial pump, which accommodates all constraints required from the integral reactor system. The main coolant pump does not have mechanical seal device because the rotor of motor and the shaft of impeller are the same one. There is no flywheel on the rotating shaft of main coolant pump so that the coastdown duration time is short when the electricity supply is cut off

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)

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)

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

Liquid metal coolants for fusion-fission hybrid system: A neutronic analysis

Energy Technology Data Exchange (ETDEWEB)

Marques, Renato V.A.; Velasquez, Carlos E.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Costa, Antonella L., E-mail: claubia@nuclear.ufmg.br [Universidade de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Barros, Graiciany P. [Comissão Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-07-01

Based on a work already published by the UFMG Nuclear Engineering Department, it was suggested to use different coolant materials in a fusion-fission system after a fuel burnup simulation, including that one used in reference work. The goal is to compare the neutron parameters, such as the effect multiplication factor and actinide amounts in transmutation layer, for each used coolant and find the best(s) coolant material(s) to be applied in the considered system. Results indicate that the lead and lead-bismuth coolant are the most suitable choices to be applied to cool the system. (author)

Main coolant pump testing at Ontario Hydro

International Nuclear Information System (INIS)

Hartlen, R.

1991-01-01

This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

The mechanisms underlying corrosion product formation and deposition in nuclear power plant circuits through the action of galvanic and thermal electromotive forces

International Nuclear Information System (INIS)

Brusakov, V.P.; Sedov, V.M.; Khitrov, Yu.A.; Brusov, K.N.; Razmashkin, N.V.; Versin, V.V.; Rybalchenko, I.L.

1983-01-01

From a theoretical standpoint, the processes of formation of corrosion products in nuclear power plant circuits, deposition of corrosion products on the circuit surfaces, formation of an equilibrium concentration of corrosion products in the coolant, and distribution of radionuclides resulting from corrosion in different parts of the circuit are considered. It is shown that the main driving forces for the mass-transfer processes in the circuits are the thermal and galvanic electromotive forces (EMF) of the microcouples. On the basis of the theoretical concepts developed the authors have obtained analytical dependences for calculating the individual stages of the process of corrosion product transfer in the circuits. The mechanisms underlying the processes which occur as a result of thermal and galvanic EMFs are considered, together with the factors influencing these processes. The results of verification of the dependences by computational methods are given and they are compared with operational data from nuclear and conventional thermal power plants and with experimental data. (author)

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.)

Breakup of jet and drops during premixing phase of fuel coolant interactions

Energy Technology Data Exchange (ETDEWEB)

Haraldsson, Haraldur Oskar

2000-05-01

During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

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

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.)

Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

National Research Council Canada - National Science Library

Can, Levent

2006-01-01

.... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

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

Physics study of Canada deuterium uranium lattice with coolant void reactivity analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Shin, Ho Cheol [Korea Hydro and Nuclear Power Central Research Institute (KHNP-CRI), Daejeon (Korea, Republic of)

2017-02-15

This study presents a coolant void reactivity analysis of Canada Deuterium Uranium (CANDU)-6 and Advanced Canada Deuterium Uranium Reactor-700 (ACR-700) fuel lattices using a Monte Carlo code. The reactivity changes when the coolant was voided were assessed in terms of the contributions of four factors and spectrum shifts. In the case of single bundle coolant voiding, the contribution of each of the four factors in the ACR-700 lattice is large in magnitude with opposite signs, and their summation becomes a negative reactivity effect in contrast to that of the CANDU-6 lattice. Unlike the coolant voiding in a single fuel bundle, the 2 x 2 checkerboard coolant voiding in the ACR-700 lattice shows a positive reactivity effect. The neutron current between the no-void and voided bundles, and the four factors of each bundle were analyzed to figure out the mechanism of the positive coolant void reactivity of the checkerboard voiding case. Through a sensitivity study of fuel enrichment, type of burnable absorber, and moderator to fuel volume ratio, a design strategy for the CANDU reactor was suggested in order to achieve a negative coolant void reactivity even for the checkerboard voiding case.

Reactor pressure vessel and reactor coolant circuit cast duplex stainless steel components contribution of the expertise for life management studies

International Nuclear Information System (INIS)

Bezdikian, Georges

2006-09-01

The life management of French Nuclear Power Plants is a major stake from an economic and a technical point of view considering the aging management assessment of the key components of the plant. The actual life evaluation is the result of prediction of life assessment from important program of expertise for the 3-loop PWR and 4-loop PWR plants in operation. To optimize the strategic policy in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of expertise considering: - the identification of degradation for different components and prediction criteria proposed; - the large database from cast reactor coolant and component removed from nuclear power plants and expertise studies to confirm the prediction; - the life evaluation of RPV with radiation surveillance program based on the expertise of irradiation capsules, it is particularly shown how the expertise is in the center of the strategic choice. The French utility has organized the life management of nuclear plant as a function of several programs of expertise of knowledge on the long term experience feedback and the maintenance program for life. This paper shows updated on RPV and reactor coolant equipment activities engaged by utility on: - periodic maintenance and volume of expertise; - Alternative maintenance actions; - Large volume of expertise and how are managed these results to predict the aging management. (author)

ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

Science.gov (United States)

Morrison, Russell H.; Holt, Mike

2006-01-01

The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

Evaluation of filtration and distillation methods for recycling automotive coolant

International Nuclear Information System (INIS)

Randall, P.M.; Gavaskar, A.R.

1992-01-01

Government regulations and high waste disposal cost of spent automotive coolant have driven the vehicle maintenance industry to explore on-site recycling. The USEPA in cooperation with the New Jersey Department of Environmental Protection (NJDEP) and the New Jersey Department of Transportation (NJDOT) evaluated two commercially available technologies that have potential for reducing the volume of spent automotive coolant. The objective of this study was to evaluate the quality of the recycled coolant, the pollution prevention potential, and the economic feasibility of the technologies

Technical findings related to Generic Issue 23: Reactor coolant pump seal failure

International Nuclear Information System (INIS)

Ruger, C.J.; Luckas, W.J. Jr.

1989-03-01

Reactor coolant pumps contain mechanical seals to limit the leakage of pressurized coolant from the reactor coolant system to the containment. These seals have the potential to leak, and a few have degraded and even failed resulting in a small break loss of coolant accident (LOCA). As a result, ''Reactor Coolant Pump Seal Failure,'' Generic Issue 23 was established. This report summarizes the findings of a technical investigation generated as part of the program to resolve this issue. These technical findings address the various fact-finding issue tasks developed for the action plan associated with the generic issue, namely background information on seal failure, evaluation of seal cooling, and mechanical- and maintenance-induced failure mechanisms. 46 refs., 15 figs., 14 tabs

Thermal hydraulic analysis of aggressive secondary cooldown in small break loss of coolant accident with total loss of high pressure safety injection

International Nuclear Information System (INIS)

Han, S. J.; Im, H. K.; Yang, J. U.

2003-01-01

Recently, Probabilistic Safety Assessment (PSA) has being applied to various fields as a basic technique of Risk-Informed Applications (RIA). To use RIA, the present study focuses on the detailed thermal hydraulic analyses for major accident sequences and success criteria to support a development of PSA model for Korea Standard Nuclear Power plant (KSNP). The primary purpose of the present study is to evaluate the success criteria of Aggressive Secondary Cooldown (ASC) in Small Break Loss Of Coolant Accident (SBLOCA) with total loss of High Pressure Safety Injection (HPSI) and to enhance the understanding of related thermal hydraulic behavior and phenomena. The accident scenario was 2 inch coldleg break LOCA without HPSI, with 1/2 Low Pressure Safety Injection (LPSI), and performing ASC limited by 55.6 .deg. C /hr (100 .deg. F/hr) cooldown rate at 15 minute after reactor trip, which successively reaches the LPSI condition for about 1.5hr after starting ASC operation with the Peak Cladding Temperature (PCT) of the hottest rod below the core damage criteria 1204.4 .deg. C (2200 .deg. F). In the present study, more relaxed success criteria than the previous PSA for KSNP could be generated under an assumption that operator should maintain the adequate ASC operation. However, it is necessary to evaluate uncertainties arisen from the related parameters of the ASC operation

Software for the diagnostic system of the secondary circuit of the Temelin nuclear power plant

International Nuclear Information System (INIS)

Drab, J.

1990-01-01

The secondary circuit of unit 1 of the Temelin nuclear power plant will be fitted with an automated diagnostic system, whose objects include the turbine and generator; feedwater pumps and their turbines; separator-reheater; condensers; low-pressure and high-pressure heaters; feedwater tank; and steam lines. The automated diagnostic system is divided into 5 subsystems, each containing a measuring unit controlled by a PC 286 computer. These computers are included in a LAN network with a PC 386 master computer. The software consists of 3 components, viz. ONSPEC for controlling the measuring unit, data evaluation and organization and for intercommunication within the LAN; diagnostic software for the diagnostic tests, of which a total of 23 are included; and communication software for transmitting the diagnostic test results to the unit control room and also for transmitting data from accurate sensors to the information computer system for technico-economic calculations. The whole system is open to future supplementing with additional software, diagnostic tests or diagnostic subsystems. (P.A.). 1 fig., 3 refs

Reactor coolant pump for a nuclear reactor

International Nuclear Information System (INIS)

Burkhardt, W.; Richter, G.

1976-01-01

An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

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

Revised Mark 22 coolant temperature coefficients

International Nuclear Information System (INIS)

Graves, W.E.

1987-01-01

Coolant temperature coefficients for the Mark 22 charge published previously are non-conservative because of the neglect of a significant mechanism which has a positive contribution to reactivity. Even after correcting for this effect, dynamic tests made on a Mark VIB charge in the early 60's suggest the results are still non-conservative. This memorandum takes both of these sources of information into account in making a best estimate of the prompt (coolant plus metal) temperature coefficient. Although no safety issues arise from this work (the overall temperature coefficient still strongly contributes to reactor stability), it is obviously desirable to use best estimates for prompt coefficients in limits and other calculations

High temperature alloys for the primary circuit of a prototype nuclear process heat plant

International Nuclear Information System (INIS)

Ennis, P.J.; Schuster, H.

1979-01-01

As part of a comprehensive materials test programme for the High Temperature Reactor Project 'Prototype Plant for Nuclear Process Heat' (PNP), high temperature alloys are being investigated for primary circuit components operating at temperatures above 750 0 C. On the basis of important material parameters, in particular corrosion behaviour and mechanical properties in primary coolant helium, the potential of candidate alloys is discussed. By comparing specific PNP materials data with the requirements of PNP and those of conventional plant, the implications for the materials programme and component design are given. (orig.)

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

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

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.

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

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Schulz, T.L.; Corletti, M.M.

1994-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit by pumping water from an in-containment refueling water storage tank during staged depressurization of the coolant circuit, the final stage including passive emergency cooling by gravity feed from the refueling water storage tank to the coolant circuit and to flood the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and avoids the final stage of depressurization with its flooding of the containment when such action is not necessary, but does not prevent the final stage when it is necessary. A high pressure makeup water storage tank coupled to the reactor coolant circuit holds makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal system can also be coupled in a loop with the refueling water supply tanks for cooling the tank. (Author)

Control of reactor coolant flow path during reactor decay heat removal

International Nuclear Information System (INIS)

Hunsbedt, A.N.

1988-01-01

This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

Procedure to determine the optimal parameters of the main primary coolant pump after compacting the FRG-1 reactor. Pt. 2. Partial structures of the procedure

International Nuclear Information System (INIS)

Pihowicz, W.

1999-01-01

On the basis of an extensive physical and technical analysis the partial structures of the procedure had been developed. They represent a logical linkage of determination elements in the form of decision and result units. The developed partial structures enable to determine the physical parameters, which characterize the primary circuit together with the compact core as well as the main primary coolant pump coming into question after compacting the core. The report also contains a discussions and a comparison of the partial structures. (orig.) [de

Development of Coolant Radioactivity Interpretation Code

International Nuclear Information System (INIS)

Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

2013-01-01

In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, Michael M.; Schulz, Terry L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path.

Nuclear reactor with makeup water assist from residual heat removal system

International Nuclear Information System (INIS)

Corletti, M.M.; Schulz, T.L.

1993-01-01

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures

Nuclear reactor with makeup water assist from residual heat removal system

Science.gov (United States)

Corletti, M.M.; Schulz, T.L.

1993-12-07

A pressurized water nuclear reactor uses its residual heat removal system to make up water in the reactor coolant circuit from an in-containment refueling water supply during staged depressurization leading up to passive emergency cooling by gravity feed from the refueling water storage tank, and flooding of the containment building. When depressurization commences due to inadvertence or a manageable leak, the residual heat removal system is activated manually and prevents flooding of the containment when such action is not necessary. Operation of the passive cooling system is not impaired. A high pressure makeup water storage tank is coupled to the reactor coolant circuit, holding makeup coolant at the operational pressure of the reactor. The staged depressurization system vents the coolant circuit to the containment, thus reducing the supply of makeup coolant. The level of makeup coolant can be sensed to trigger opening of successive depressurization conduits. The residual heat removal pumps move water from the refueling water storage tank into the coolant circuit as the coolant circuit is depressurized, preventing reaching the final depressurization stage unless the makeup coolant level continues to drop. The residual heat removal system can also be coupled in a loop with the refueling water supply tank, for an auxiliary heat removal path. 2 figures.

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.)

Transient behaviour of main coolant pump in nuclear power plants

International Nuclear Information System (INIS)

Delja, A.

1986-01-01

A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

Vertical reactor coolant pump instabilities

International Nuclear Information System (INIS)

Jones, R.M.

1985-01-01

The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corractive measures taken are also described

Study of transitory regimes in hydraulic cooling circuits

International Nuclear Information System (INIS)

Bonnin, Jacques; Fanelli, Michel.

1975-01-01

The problem of transient regimes operated voluntary or not in hydraulic circuits is posed and the risks they cause are shown. As for the case of coolant flow loss the various methods for studying the problem are examined: numerical simulation (explicit and implicit), physical model, on-site testing. The numerical methods that not yet fully satisfying or economic, are still very badly representative for hollow closures. Physical models, expensive in the case of a first facility, are not still fully representative (inconsistent similitudes, difficulties in pump picturing). Site test recordings are often a trouble for exploitation and always limited to nondestructive tests. Comparison between the three methods, already satisfying, will have to be improved to allow remedies to the over pressures due to the transients to be developed [fr

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

Preliminary design of reactor coolant pump canned motor for AC600

International Nuclear Information System (INIS)

Deng Shaowen

1998-01-01

The reactor coolant pump canned motor of AC600 PWR is the kind of shielded motors with high moment of inertia, high reliability, high efficiency and nice starting performance. The author briefly presents the main feature, design criterion and technical requirements, preliminary design, computation results and analysis of performance of AC600 reactor coolant pump canned motor, and proposes some problems to be solved for study and design of AC600 reactor coolant pump canned motor

Multivariable robust control of an integrated nuclear power reactor

Directory of Open Access Journals (Sweden)

A. Etchepareborda

2002-12-01

Full Text Available The design of the main control system of the CAREM nuclear power plant is presented. This plant is an inherently safe low-power nuclear reactor with natural convection on the primary coolant circuit and is self-pressurized with a steam dome on the top of the pressure vessel (PV. It is an integrated reactor as the whole primary coolant circuit is within the PV. The primary circuit transports the heat to the secondary circuit through once-through steam generators (SG. There is a feedwater valve at the inlet of the SG and a turbine valve at the outlet of the SG. The manipulated variables are the aperture of these valves and the reactivity of the control rods. The control target is to regulate the primary and secondary pressures and to monitor steam flow reference ramps on a range of nominal flow from 100% to 40%. The requirements for the control system are robust stability, low-order simple controllers and transient/permanent error bounding. The controller design is based on a detailed RETRAN plant model, from which linear perturbed open-loop dynamic models at different powers are identified. Two low-order nominal models with their associated uncertainties are chosen for two different power ranges. Robust controllers with acceptable performances are designed for each range. Numerical optimization based on the loop-shaping method is used for the controller design. The designed controllers are implemented in the RETRAN model and tested in simulations achieving successful results.

Evaluation of primary coolant leaks and assessment of detection methods

International Nuclear Information System (INIS)

Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

1986-01-01

A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

Evaluation of primary coolant leaks and assessment of detection methods

International Nuclear Information System (INIS)

Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

1984-11-01

A review of French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all occurred leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by the compliance with the criteria defined in the operating technical specifications

High temperature filtration of radioactivable corrosion products in the primary circuit of PWR type reactors

International Nuclear Information System (INIS)

Dolle, L.

1976-01-01

A effective limitation to the deposition of radioactive corrosion products in the core of a reactor at power operation, is to be obtained by filtering the water of the primary circuit at a flow rate upper than 1% of the coolant flow rate. However, in view of accounting for more important release of corrosion products during the reactor start-up and also for some possible variations in the efficiency of the system, it is better that the flow rate to be treated by the cleaning circuit is stated at 5%. Filtration must be effected at the temperature of the primary circuit and preferably on each loop. To this end, the feasibility of electromagnetic filtration or filtration through a deep bed of granulated graphite has been studied. The on-loop tests effected on each filter gave efficiencies and yields respectively upper than 90% and 99% for magnetite and ferrite particles in suspension in water at 250 deg C. Such results confirm the interest lying in high temperature filtration and lead to envisage its application to reactors [fr

Testing installation for a steam generator

International Nuclear Information System (INIS)

Dubourg, M.

1985-01-01

The invention proposes a testing installation for a steam generator associated to a boiler, comprising a testing exchanger connected to a feeding circuit in secondary fluid and to a circuit to release the steam produced, and comprising a heating-tube bundle connected to a closed circuit of circulation of a primary coolant at the same temperature and at the pressure than the primary fluid. The heating-tube bundle of the testing exchanger has the same height than the primary bundle of the steam generator and the testing exchanger is at the same level and near the steam generator and is fed by the same secondary fluid such as it is subject to the same operation phases during a long period. The in - vention applies, more particularly, to the steam generators of pressurized water nuclear power plants [fr

The solid coolant and prospects of its use in innovative reactors

International Nuclear Information System (INIS)

Dmitriev, A.M.; Deniskin, V.P.

2010-01-01

The progress of nuclear power demands consideration and development of innovative projects of the reactors having the increased level of safety due to their immanent properties allowing to provide high parameters. One of interesting and perspective offers is the use of a solid substance as a coolant. Use of the solid coolant of a nuclear reactor core has significant advantages among which an opportunity of movement of the coolant in the core under action of gravities and absence of necessity to have superfluous pressure in the jacket, that in turn means small metal consumption of construction, decrease in risk of emergency and its consequences. Cooling of the core with the help of solid substance is possible at performance of the certain conditions connected to features of the solid coolant. The major requirements are: the uniform continuous movement and minimal fluctuation of its density on every site of the core; high mechanical durability and wear resistance of particles; as well as good parameters of heat exchange, i.e. high heat conductivity and thermal capacity of the coolant material at the core operating conditions

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

Physical model and calculation code for fuel coolant interactions

International Nuclear Information System (INIS)

Goldammer, H.; Kottowski, H.

1976-01-01

A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

LOFT advanced densitometer for nuclear loss-of-coolant experiments

International Nuclear Information System (INIS)

Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

1979-01-01

A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

Assistance in chemistry and chemical processes related to primary, secondary and ancillary systems of nuclear power plants

International Nuclear Information System (INIS)

Chocron, Mauricio A.; Becquart, Elena T.; Iglesias, Alberto M.; La Gamma, Ana M.; Villegas, Marina

2003-01-01

Argentina is currently running two nuclear power plants: Atucha I (CNA I) and Embalse (CNE) operated by Nucleoelectrica Argentina (NASA) whereas the National Atomic Energy Commission (CNEA), among other activities, is responsible for research and development in the nuclear field, operates research reactors and carries out projects related to them. In particular, the Reactor Chemistry Section personnel (currently part of the Chemistry Dept.) has been working on the field of reactor water chemistry for more than 25 years, on research and support to the NPPs chemistry department. Though the most relevant tasks have been connected to primary and secondary circuits chemistry, ancillary systems show along the time unexpected problems or feasible improvements originated in the undergoing operating time as well as in phenomena not foreseen by the constructors. In the present paper are presented the tasks performed in relation to the following systems of Embalse NPP: 1) Heavy water upgrade column preliminary water treatment; 2) Liquid waste system preliminary water treatment; and 3) Primary heat transport system coolant crud composition. (author)

Corrosion products in the primary circuits of PWRs

International Nuclear Information System (INIS)

Darras, R.

1983-01-01

The characteristics of PWR primary circuits are recalled, particularly the chemical specifications of the medium and the various materials used (austenitic steel, nickel alloys, cobalt-based alloys and zirconium alloys). The behaviour of these materials as regards general corrosion in nominal and transient conditions is then outlined briefly, special emphasis being laid on the effect of the determining parameters on the quantity of corrosion products formed. The release of the latter into the primary coolant is caused by two main processes: solubilization and erosion. Particular attention was given therefore to the laws governing the solubility of the oxides involved, especially as a function of temperature and pH. Erosion, or release in the form of solid particles, is relatively severe during transient events. As these corrosion products are then carried through all circuits, they cause deposits to form in favourable places on the walls as a result either of precipitation of soluble species or of sedimentation followed by consolidation of suspended particles. The presence of corrosion products in the primary circuits creates a particular impact since they become radioactive as they pass through the core and especially when they remain in it in the form of deposits; as a result, the products are capable of contaminating the entire system. Finally, although long-term reliability is obviously an essential condition for materials developed, attention must also be given to problems associated with a build-up of corrosion products in the cooling circuits and efforts made to minimize them. To that end, a number of precautions are recommended, and various remedies can be applied: selecting materials which are not readily activated, keeping structures clean, purifying fluids properly, restricting solubilization and precipitation, and perhaps, periodic decontamination. (author)

Development for LMR coolant technology - Development of a submersible-in-pool electromagnetic pump

Energy Technology Data Exchange (ETDEWEB)

Hong, Sang Hee; Kim, Hee Reyoung; Lee, Sang Don; Seo, Chun Ho [Seoul National University, Seoul (Korea, Republic of); Cho, Su Won [Kyungki University, Suwon (Korea, Republic of)

1995-08-01

The conceptual and detailed designs of an annular linear induction electromagnetic pump of small scale submersible-in-pool type are performed for the purpose of domestic development of the pumps used for the high-temperature natrium coolant transportation in liquid metal reactors. The pump drawings for and input power of 1,100 VA, an input frequency of 17 Hz, a maximum flowrate of 60 l/min and a maximum operation temperature of 600 deg C are obtained from the optimum design analyses by solving MHD and equivalent circuit equations. The characteristics of pump materials in the high temperature and neutron irradiation environment are reflected in designing the pump, and theoretical analyses for improving the pump performance and efficiency are tried through calculations of magnetic flux and temperature distributions inside the pump. The present project contributes to the further design of engineering proto-type electromagnetic pump with higher capacity and the development of liquid metal reactor with innovative simplicity. 44 refs., 4 tabs., 33 figs. (author)

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.)

Nuclear power plant V-1

International Nuclear Information System (INIS)

1998-01-01

The nuclear power plant Bohunice V -1 is briefly described. This NPP consists from two reactor units. Their main time characteristics are (Reactor Unit 1, Reactor Unit 2): beginning of construction - 24 April 1972; first controlled reactor power - 27 November 1978, 15 March 1980; connection to the grid - 17 December 1978, 26 March 1980; commercial operation - 1 April 1980, 7 January 1981. This leaflet contains: NPP V-1 construction; Major technological equipment (Primary circuit: Nuclear reactor [WWER 440 V230 type reactor];Steam generator; Reactor Coolant Pumps; Primary Circuit Auxiliary Systems. Secondary circuit: Turbine generators, Nuclear power plant electrical equipment; power plant control) and technical data

Modeling of Hydrodynamic Processes at a Large Leak of Water into Sodium in the Fast Reactor Coolant Circuit

Directory of Open Access Journals (Sweden)

Sergey Perevoznikov

2016-10-01

Full Text Available In this paper, we describe a physicomathematical model of the processes that occur in a sodium circuit with a variable flow cross-section in the case of a water leak into sodium. The application area for this technique includes the possibility of analyzing consequences of this leak as applied to sodium–water steam generators in fast neutron reactors. Hydrodynamic processes that occur in sodium circuits in the event of a water leak are described within the framework of a one-dimensional thermally nonequilibrium three-component gas–liquid flow model (sodium–hydrogen–sodium hydroxide. Consideration is given to the results of a mathematical modeling of experiments involving steam injection into the sodium loop of a circulation test facility. That was done by means of the computer code in which the proposed model had been implemented.

Discrete element method study of fuel relocation and dispersal during loss-of-coolant accidents

Science.gov (United States)

Govers, K.; Verwerft, M.

2016-09-01

The fuel fragmentation, relocation and dispersal (FFRD) during LOCA transients today retain the attention of the nuclear safety community. The fine fragmentation observed at high burnup may, indeed, affect the Emergency Core Cooling System performance: accumulation of fuel debris in the cladding ballooned zone leads to a redistribution of the temperature profile, while dispersal of debris might lead to coolant blockage or to debris circulation through the primary circuit. This work presents a contribution, by discrete element method, towards a mechanistic description of the various stages of FFRD. The fuel fragments are described as a set of interacting particles, behaving as a granular medium. The model shows qualitative and quantitative agreement with experimental observations, such as the packing efficiency in the balloon, which is shown to stabilize at about 55%. The model is then applied to study fuel dispersal, for which experimental parametric studies are both difficult and expensive.

Breakup of jet and drops during premixing phase of fuel coolant interactions

International Nuclear Information System (INIS)

Haraldsson, Haraldur Oskar

2000-05-01

During the course of a hypothetical severe accident in a light water reactor, molten liquid may be introduced into a volatile coolant, which, under certain conditions, results in explosive interactions. Such fuel-coolant interactions (FCI) are characterised by an initial pre-mixing phase during which the molten liquid, metallic or oxidic in nature, undergoes a breakup (fragmentation) process which significantly increase the area available for melt-coolant contact, and thus energy transfer. Although substantial progress in the understanding of phenomenology of the FCI events has been achieved in recent years, there remain uncertainties in describing the primary and secondary breakup processes. The focus of this work is on the melt jet and drop breakup during the premixing phase of FCI. The objectives are to gain insight into the premixing phase of the FCI phenomena, to determine what fraction of the melt fragments and determine the size distribution. The approach is to perform experiments with various simulant materials, at different scales, different conditions and with variation of controlling parameters affecting jet and drop breakup processes. The analysis approach is to investigate processes at different level of detail and complexity to understand the physics, to rationalise experimental results and to develop and validate models. In the first chapter a brief introduction and review of the status of the FCI phenomena is performed. A review of previous and current experimental projects is performed. The status of the experimental projects and major findings are outlined. The first part of the second chapter deals with experimental investigation of jet breakup. Two series of experiments were performed with low and high temperature jets. The low temperature experiments employed cerrobend-70 as jet liquid. A systematic investigation of thermal hydraulic conditions and melt physical properties on the jet fragmentation and particle debris characteristics was

Operational behaviour of a reactor normal operation and disturbances

International Nuclear Information System (INIS)

Geyer, K.H.

1982-01-01

During normal operation, the following topics are dealt with: primary and secondary coolant circuits - full load operation - start-up and shutdown - steady state part load diagramm. During disturbances and incidents, the following procedures are discussed: identification and detection of the events - automatic actions - manual actions of the operator - provided indications - explanation of actuated systems - basic information of reactor protection system. (RW)

Multi-state reliability for coolant pump based on dependent competitive failure model

International Nuclear Information System (INIS)

Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

2013-01-01

By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

Effect of spacer grid mixing vanes on coolant outlet temperature distribution

Energy Technology Data Exchange (ETDEWEB)

Raemae, Tommi; Lahtinen, Tuukka; Brandt, Tellervo; Toppila, Timo [Fortum Power and Heat, Fortum (Finland). Nuclear Competence Center

2012-08-15

In Loviisa VVER-440-type NPP the coolant outlet temperature of the hot subchannel is constantly monitored during the operation. According to the authority requirement the maximum subchannel outlet temperature must not exceed the saturation temperature. Coolant temperature distribution inside the fuel assembly is affected by the efficiency of the coolant mixing. In order to enhance the coolant mixing the fuel manufacturer is introducing the additional mixing vanes on the fuel bundle spacer grids. In the paper the effect of the different mixing vane modifications is studied with computational fluid dynamics (CFD) simulation. Goal of the modelling is to find vane modifications with which sufficient mixing is reached with acceptable increase in the spacer grid pressure loss. The results of the studies are discussed in the paper. (orig.)

Radiological characterization of the main circulation circuit of the 1st unit at Ignalina NPP

International Nuclear Information System (INIS)

Poskas, P.; Zujus, R.; Brazauskaite, A. and others

2003-01-01

A preliminary assessment of the main circulation circuit (MCC) equipment contamination at the reactor final shut-down and analysis of the contamination variation with time are presented. In order to perform this assessment, data on MCC element characteristics, the coolant characteristics, the system operational conditions, etc. were collected. The modified computer code LLWAA-DECOM (Belgium) was used to determine contamination of the system elements. As a result, nuclide vectors and surface dose rates of deposits for every MCC element were determined. The modeling results were compared to the measurement results of the dose rates. (author)

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.

Coolant voiding analysis following SGTR for an HLMC reactor

International Nuclear Information System (INIS)

Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

2000-01-01

Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

Analysis of the behaviour of pressure and temperature of the containment of a PWR reactor, submitted to a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Silva, D.E. da; Arrieta, L.A.J.; Costa, J.R.; Camargo, C.; Santos, C.M. dos; Rochedo, E.R.R.

1979-12-01

The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary circuit. The scope of the study was directed to verify the Final Safety Analysis Report (FSAR) results for the integrity of the metalic containment of the Angra I power plant. The highest containment pressure peak for this unit is expected for a break in the suction line of one of the main pumps of the primary coolant. Using the same input data, our results are very similar to those presented in the FSAR which shows a reasonable equivalence between the two analytical models. Using as input data the results of a previous LOCA study at CNEN, which yields to more conservative boundary conditions than those presented by the FSAR, the pressure and temperature peak values determined by our model are quite larger than those presented by the cited Safety Report. (author) [pt

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.)

Labelling Of Coolant Flow Anomaly Using Fractal Structure

International Nuclear Information System (INIS)

Djainal, Djen Djen

1996-01-01

This research deals with the instrumentation of the detection and characterization of vertical two-phase flow coolant. This type of work is particularly intended to find alternative method for the detection and identification of noise in vertical two-phase flow in a nuclear reactor environment. Various new methods have been introduced in the past few years, an attempt to developed an objective indicator off low patterns. One of new method is Fractal analysis which can complement conventional methods in the description of highly irregular fluctuations. In the present work, Fractal analysis was applied to analyze simulated boiling coolant signal. This simulated signals were built by sum random elements in small subchannels of the coolant channel. Two modes are defined and both are characterized by their void fractions. In the case of uni modal -PDF signals, the difference between these modes is relatively small. On other hand, bimodal -PDF signals have relative large range. In this research, Fractal dimension can indicate the characters of that signals simulation

Sloshing of coolant in a seismically isolated reactor

International Nuclear Information System (INIS)

Wu, T.S.; Guildys, J.; Seidensticker, R.W.

1988-01-01

During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

Flow accelerated corrosion and its control measures for the secondary circuit pipelines in Indian nuclear power plants

International Nuclear Information System (INIS)

Kain, Vivekanand; Roychowdhury, S.; Mathew, Thomas; Bhandakkar, Atul

2008-01-01

A plain carbon steel feeder pipeline in the secondary circuit failed downstream of a flow measurement device (orifice meter) during operation at nuclear power plant. A detailed failure analysis done on the failed pipeline is described in this paper. The results established the fine surface pattern of 'Horseshoe pits' at the affected regions. X-ray diffraction analysis on the samples far from the failed regions showed presence of magnetite but on the sample from the failed region showed peaks due to base metal only, indicating dissolution of the oxide. Thickness profiling of the pipeline indicated reduction of thickness from the design 7.62 mm to a minimum of 0.4-1.4 mm at the location of the failure. These observations are characteristic of single phase flow accelerated corrosion. This paper details the extent of flow accelerated corrosion in various Indian power plants and the remedial measures for replacement and possible design and water chemistry changes to combat it

Factors influencing the renal arterial Doppler waveform: a simulation study using an electrical circuit model (secondary publication)

Energy Technology Data Exchange (ETDEWEB)

Sung, Chang Kyu [Dept. of Radiology, SMG-SNU Boramae Medical Center, Seoul National University College of Medicine, Seoul (Korea, Republic of); Han, Bong Soo [Dept. of Radiological Science, College of Health Science, Yonsei University, Wonju (Korea, Republic of); Kim, Seung Hyup [Dept. of Radiology, Institute of Radiation Medicine, Seoul National University Hospital, Seoul National University College of Medicine, Seoul (Korea, Republic of)

2016-01-15

The goal of this study was to evaluate the effect of vascular compliance, resistance, and pulse rate on the resistive index (RI) by using an electrical circuit model to simulate renal blood flow. In order to analyze the renal arterial Doppler waveform, we modeled the renal blood-flow circuit with an equivalent simple electrical circuit containing resistance, inductance, and capacitance. The relationships among the impedance, resistance, and compliance of the circuit were derived from well-known equations, including Kirchhoff’s current law for alternating current circuits. Simulated velocity-time profiles for pulsatile flow were generated using Mathematica (Wolfram Research) and the influence of resistance, compliance, and pulse rate on waveforms and the RI was evaluated. Resistance and compliance were found to alter the waveforms independently. The impedance of the circuit increased with increasing proximal compliance, proximal resistance, and distal resistance. The impedance decreased with increasing distal compliance. The RI of the circuit decreased with increasing proximal compliance and resistance. The RI increased with increasing distal compliance and resistance. No positive correlation between impedance and the RI was found. Pulse rate was found to be an extrinsic factor that also influenced the RI. This simulation study using an electrical circuit model led to a better understanding of the renal arterial Doppler waveform and the RI, which may be useful for interpreting Doppler findings in various clinical settings.

Factors influencing the renal arterial Doppler waveform: a simulation study using an electrical circuit model (secondary publication)

International Nuclear Information System (INIS)

Sung, Chang Kyu; Han, Bong Soo; Kim, Seung Hyup

2016-01-01

The goal of this study was to evaluate the effect of vascular compliance, resistance, and pulse rate on the resistive index (RI) by using an electrical circuit model to simulate renal blood flow. In order to analyze the renal arterial Doppler waveform, we modeled the renal blood-flow circuit with an equivalent simple electrical circuit containing resistance, inductance, and capacitance. The relationships among the impedance, resistance, and compliance of the circuit were derived from well-known equations, including Kirchhoff’s current law for alternating current circuits. Simulated velocity-time profiles for pulsatile flow were generated using Mathematica (Wolfram Research) and the influence of resistance, compliance, and pulse rate on waveforms and the RI was evaluated. Resistance and compliance were found to alter the waveforms independently. The impedance of the circuit increased with increasing proximal compliance, proximal resistance, and distal resistance. The impedance decreased with increasing distal compliance. The RI of the circuit decreased with increasing proximal compliance and resistance. The RI increased with increasing distal compliance and resistance. No positive correlation between impedance and the RI was found. Pulse rate was found to be an extrinsic factor that also influenced the RI. This simulation study using an electrical circuit model led to a better understanding of the renal arterial Doppler waveform and the RI, which may be useful for interpreting Doppler findings in various clinical settings

Utilization of DRUFAN 01/MOD 02 computer code for the depressurization phase analysis of a postulated loss of coolant accident in Angra 2/3 Nuclear Power Plants

International Nuclear Information System (INIS)

Austregesilo Filho, H.; Figueiredo, P.J.M.

1985-08-01

The DRUFAN 01/Mod 2 developed by Gesellschaft fur Reaktorsicherheit (GRS) mbh to simulate thermohydraulic behavior of the primary circuit of PWR reactors, during the despressurization phase and initial refilling phase of loss of coolant accidents by great ruptures, is presented. The program simulates the system to be analysed by control volumes-concentrated parameters model - and it is based on numerical solution of conservation equations for mass of water, mass of vapor, quantities of motion and energy, and on the control volume homogeneity hypothesis. The possibilities of thermodynamic disequilibrium, determining mass transfer between liquid and vapor phases assuming that one saturated phase, are considered. The process of computer code implantation in the Honeywell Bull 64 DPS 7 system at CNEN, the modifications done into the program and the application to the despressurization phase analysis of a loss of coolant accident at Angra-2 and Angra-3 reactors are considered. (M.C.K.) [pt

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.)

Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

International Nuclear Information System (INIS)

Luo Bangqi

2012-01-01

The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

Corrosion of alloy 800 in PHWR primary and secondary conditions

International Nuclear Information System (INIS)

Maroto, A.J.G.; Blesa, M.A.; Villegas, M.; Olmedo, A.M.; Bordoni, R.; Alvarez, M.G.; Sainz, R.

1998-01-01

A hot leg section of a steam generator tubing was removed for destructive examination from one of the steam generators (SG) of the Embalse Nuclear Power Plant. The tube material is Alloy 800 and carbon steel is the tube support plate material. Samples of the deposits were taken at the first tube support plate and at the top, mid-height and bottom of the sludge pile. Transverse sections were taken at several locations along the tube length measuring the oxide thicknesses and studying the morphology of the oxide layer by scanning electron microscopy on the primary and secondary side at each location. Deposit layers on the outer tube surface revealed iron as major component and the presence of calcium, phosphorous, zinc and manganese. The oxide scale thickness at the secondary side in the open area was around 22 to 30 μm. The oxide thickness grown under isothermal conditions on the corrosion test samples installed in the autoclaves facilities of the primary circuit of the plant was measured and compared with that found on the inner surface of the examined tube section. The oxide thickness of the test samples was around 1-2 μm showing the influence of the deposition of corrosion products from the coolant. Deposition and precipitation of oxide was also found in the actual tube, where the common feature was the irregularity of the oxide layer on the primary side and thicknesses values in the range 4 to 10 μm were measured. The autoclave tests and SG tubing examination permit to compare the influence of materials and of operating (flow rate, isothermal vs non-isothermal) conditions on corrosion and deposition. (author)

Standardized sampling system for reactor coolants

International Nuclear Information System (INIS)

Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

1982-09-01

A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

Automatic coolant flow control device for a nuclear reactor assembly

Science.gov (United States)

Hutter, Ernest

1986-01-01

A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

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.

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

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.

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.

Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

Energy Technology Data Exchange (ETDEWEB)

Robbe, M.F

2003-12-01

Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

Primary system hydraulic characteristics after modification of reactor coolant pumps' impeller wheels at Bohunice NPP executed in 2012 and 2013

International Nuclear Information System (INIS)

Hermansky, Jozef; Zavodsky, Martin

2014-01-01

A coolant flow through the reactor is usually determined after annual outages at Slovak NPP (VVER 440) in two distinct ways. First method is determination on the basis of the secondary system parameters - measurement of thermal balances. The value achieved by this method is used as the input parameter in the Table of allowed reactor operation modes. The second method draws from the primary system parameters - measurement of primary system hydraulic characteristics. Flow nozzles used for the measurement of feed water flow behind high pressure heaters were replaced at both Bohunice Units during outages in 2008. The feed water flow behind high pressure heaters is one of the main parameters used for the determination of coolant flow through the reactor by the first method. Compared to the measurement executed during previous fuel cycles, the calculated coolant flow through the reactor decreased considerably after the change of flow nozzles. The imaginary change of coolant flow through the reactor at Unit 3 was -1,6 %; and at Unit 4 -2,6 %. This change was not proved by the parallel measurement of primary system hydraulic characteristics. Later it was found out that the original flow nozzles used for 25 years were substantially deposited (original inner diameter of the nozzles was reduced by about 0,6-0,9 mm). Therefore feed water flow measurement was untrustworthy within the recent years. On the findings stated above, Bohunice NPP has decided to increase coolant flow through the reactor by changing the reactor coolant pumps impeller wheels. The modification of impellers wheels is planned within years 2012 to 2014. During the outages in 2013 two impeller wheels were replaced at both units. Nowadays Unit 4 is operated with all 6 new impeller wheels and Unit 3 with four new impeller wheels. Modification of last two impeller wheels at Unit 3 will be performed during the outage in 2014. On account of impeller wheels modification, non-standard measurement of PS hydraulic

Fatigue check of nuclear safety class 1 reactor coolant pipe

International Nuclear Information System (INIS)

Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong

2015-01-01

Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)

Method for controlling a coolant liquid surface of cooling system instruments in an atomic power plant

International Nuclear Information System (INIS)

Monta, Kazuo.

1974-01-01

Object: To prevent coolant inventory within a cooling system loop in an atomic power plant from being varied depending on loads thereby relieving restriction of varied speed of coolant flow rate to lowering of a liquid surface due to short in coolant. Structure: Instruments such as a superheater, an evaporator, and the like, which constitute a cooling system loop in an atomic power plant, have a plurality of free liquid surface of coolant. Portions whose liquid surface is controlled and portions whose liquid surface is varied are adjusted in cross-sectional area so that the sum total of variation in coolant inventory in an instrument such as a superheater provided with an annulus portion in the center thereof and an inner cylindrical portion and a down-comer in the side thereof comes equal to that of variation in coolant inventory in an instrument such as an evaporator similar to the superheater. which is provided with an overflow pipe in its inner cylindrical portion or down-comer, thereby minimizing variation in coolant inventory of the entire coolant due to loads thus minimizing variation in varied speed of the coolant. (Kamimura, M.)

Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

2014-10-15

A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

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)

Design and development of remotely operated coolant channel cutting machine

International Nuclear Information System (INIS)

Suthar, R.L.; Sinha, A.K.; Srikrishnamurty, G.

1994-01-01

One of the coolant tubes of Narora Atomic Power Station (NAPS) reactor needs to be removed. To remove a coolant tube, four cutting operations, (liner tube cutting, end-fitting cutting, machining of seal weld of bellow ring and finally coolant tube cutting) are required to be carried out. A remotely operated cutting machine to carry out all these operations has been designed and developed by Central Workshops. This machine is able to cut at the exact location because of numerically controlled axial and radial travel of tool. Only by changing the tool head and tool holder, same machine can be used for various types of cutting/machining operations. This report details the design, manufacture, assembly and testing work done on the machine. (author). 4 figs

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.

Dryout delay in loss-of-coolant incidents in nuclear power plants

International Nuclear Information System (INIS)

Belda, W.

1975-01-01

The maximum credible accident (MCA) as a result of a fault in the system is assumed to be the rupture of a pipe in the primary circuit. During the outflow process following the rupture - called blowdown - it is possible that the internals of a reactor pressure vessel are exposed to extreme mechanical and thermal stresses. The fuel rods in the core, the Zircaloy cladding tubes of which can be heated up by lack of coolant to inadmissibly high temperatures, are particularly at risk. In case of the cladding tubes being damaged, radioactive substances are released. If they escape from the outer containment, this would lead to pressures on the immediate and more distant vicinity of the nuclear pover plant. In order to eliminate the factors of uncertainty when calculating the overall blowdown process in advance, it is necessary to have a relationship valid for the instationary circumstances to work out the burnout delay which is of decisive importance for the post-incident cooling phase of the reactor. The aim of this investigation, therefore, is to develop, with the aid of a suitable model, a method of calculating the burnout delay. (orig./TK) [de

Integrated Fuel-Coolant Interaction (IFCI 6.0) code

International Nuclear Information System (INIS)

Davis, F.J.; Young, M.F.

1994-04-01

The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

Probabilistic analyses of failure in reactor coolant piping

International Nuclear Information System (INIS)

Holman, G.S.

1984-01-01

LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

Sodium as a reactor coolant

International Nuclear Information System (INIS)

Cesar, S.B.G.

1989-01-01

This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

Nuclear power plant V-1

International Nuclear Information System (INIS)

1999-01-01

In this leaflet the short history of commissioning of Bohunice V-1 NPP is reviewed (beginning of construction 24 April 1972; First controlled reactor power, Reactor Unit 1 (RU1): 27 November 1978, Reactor Unit 2 (RU2): 15 March 1980; Connection to the grid: RU1 17 December 1978, RU2 26 March 1980; Commercial operation: RU1 1 April 1980, RU2 7 January 1981. The scheme of the nuclear reactor WWER 440/V230 is depicted. The major technological equipment (primary circuit, nuclear reactor, steam generators, reactor coolant pumps, primary circuit auxiliary systems, secondary circuit, turbine generators, NPP electrical equipment, and power plant control) are described. Technical data of the Bohunice V-1 NPP are presented

Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

International Nuclear Information System (INIS)

Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

2014-01-01

The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

Solution of multiple circuits of steam cycle HTR system

International Nuclear Information System (INIS)

Li, Fu; Wang, Dengying; Hao, Chen; Zheng, Yanhua

2014-01-01

In order to analyze the dynamic operation performance and safety characteristics of the steam cycle high temperature gas cooled reactor (HTR) systems, it is necessary to find the solution of the whole HTR systems with all coupled circuits, including the primary circuit, the secondary circuit, and the residual heat removal system (RHRS). Considering that those circuits have their own individual fluidity and characteristics, some existing code packages for independent circuits themselves have been developed, for example THEMRIX and TINTE code for the primary circuit of the pebble bed reactor, BLAST for once through steam generator. To solve the coupled steam cycle HTR systems, a feasible way is to develop coupling method to integrate these independent code packages. This paper presents several coupling methods, e.g. the equivalent component method between the primary circuit and steam generator which reflect the close coupling relationship, the overlapping domain decomposition method between the primary circuit and the passive RHRS which reflects the loose coupling relationship. Through this way, the whole steam cycle HTR system with multiple circuits can be easily and efficiently solved by integration of several existing code packages. Based on this methodology, a code package TINTE–BLAST–RHRS was developed. Using this code package, some operation performance of HTR–PM was analyzed, such as the start-up process of the plant, and the depressurized loss of forced cooling accident when different number of residual heat removal trains is operated

Performance of Helical Coil Heat Recovery Exchanger using Nanofluid as Coolant

Directory of Open Access Journals (Sweden)

Navid Bozorgan

2015-07-01

Full Text Available Nanofluids are expected to be a promising coolant condidate in chemical processes for heat transfer system size reduction. This paper focuses on reducing the number of turns in a helical coil heat recovery exchanger with a given heat exchange capacity in a biomass heating plant using γ-Al2O3/n-decane nanofluid as coolant. The nanofluid flows through the tubes and the hot n-hexane flows through the shell. The numerical results show that using nanofluid as coolant in a helical coil heat exchanger can reduce the manufacturing cost of the heat exchanger and pumping power by reducing the number of turns of the coil.

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

Simulation of a loss of coolant accident

International Nuclear Information System (INIS)

1987-06-01

An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

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

Effect of ribbed and smooth coolant cross-flow channel on film cooling

International Nuclear Information System (INIS)

Peng, Wei; Sun, Xiaokai; Jiang, Peixue; Wang, Jie

2017-01-01

Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

Effect of ribbed and smooth coolant cross-flow channel on film cooling

Energy Technology Data Exchange (ETDEWEB)

Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

2017-05-15

Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

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

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

Coolant void effect investigation - case of a na-cooled fast reactor

International Nuclear Information System (INIS)

Glinatsis, G.; Gugiu, D.

2013-01-01

In the frame of the last EURATOM-FP7 Program, a large sized Sodium-cooled FR (SFR) has been studied. Mixed carbides fuel (U, Pu)C has been adopted for the backup core solution and important work has been also performed in order to obtain an ''optimised'' backup configuration ''close'' to the reference one, which is fueled by mixed oxides fuel (U, Pu)Ox. The peculiarity of both core designs (the reference configuration and the optimised backup configuration) is the adoption of a 60 cm Plenum zone in the upper part of each fuel assembly (FA), that is filled by coolant, in order to mitigate (when emptied) the core positive coolant void effect. This paper presents some results of a detailed study of the coolant void effect for the above SFR with mixed carbides core. Many aspects, like geometric heterogeneity, the burnup state, the operating conditions, etc., have been taken into consideration in order to obtain information about the ''propagation'' and the behaviour of the coolant void effect itself. The performed study investigates also the coolant void effect consequences on some reactivity coefficients, which are important for a safe behaviour of the reactor. The investigation consisted in the steady state simulations of the reactor on different operating conditions in Monte Carlo approach. (authors)

Application of surface science to the study of the corrosion of PWR primary circuit materials

International Nuclear Information System (INIS)

Harris, S.J.

1989-04-01

This thesis describes a study of the corrosion and oxidation of PWR primary circuit materials using surface sensitive spectroscopic techniques. An X-ray photoemission spectroscopy (XPS) study of a number of mixed oxides of known composition is described and the information obtained is related to XPS measurements made on the surface of iron and nickel based alloys oxidised under controlled conditions. A secondary ion mass spectroscopy (SIMA) study on these mixed transition metal oxides is also described. The gaseous oxidation of stainless steel 3041 and Inconel-690 is examined. Both alloys were oxidised at 600K in air with the composition of the oxide films formed studied by a range of surface spectroscopic methods. Further experimental work was performed on Inconel-690 to examine the effects of surface pretreatment and the effects of low oxygen partial pressures on the formation of oxide films at 600 K. The incorporation of the radionuclide, cobalt-60, into the oxide films formed on structural components of a PWR, result in the build up of radiation fields. A method of pretreating the surface of the alloy stainless steel 3041, in order to reduce the level of cobalt adsorbed into the oxide film formed under simulated primary coolant conditions is examined and contrasts with treatments which have been developed to release cobalt adsorbed in existing oxide layers under reactor conditions are discussed. (author)

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