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Sample records for kartini water coolant

  1. Study on the effect of moderator density reactivity for Kartini reactor

    International Nuclear Information System (INIS)

    Budi Rohman; Widarto

    2009-01-01

    One of important characteristics of water-cooled reactors is the change of reactivity due to change in the density of coolant or moderator. This parameter generally has negative value and it has significant role in preventing the excursion of power during operation. Many thermal-hydraulic codes for nuclear reactors require this parameter as the input to account for reactivity feedback due to increase in moderator voids and the subsequent decrease in moderator density during operation. Kartini reactor is cooled and moderated by water, therefore, it is essential to study the effect of the change in moderator density as well as to determine the value of void or moderator density reactivity coefficient in order to characterize its behavior resulting from the presence of vapor or change of moderator density during operation. Analysis by MCNP code shows that the reactivity of core is decreasing with the decrease in moderator density. The analysis estimates the void or moderator density reactivity coefficient for Kartini Reactor to be -2.17×10-4 Δρ/ % void . (author)

  2. Safety Evaluation of Kartini Reactor Based on Instrumentation System Design

    International Nuclear Information System (INIS)

    Tjipta Suhaemi; Djen Djen Dj; Itjeu K; Johnny S; Setyono

    2003-01-01

    The safety of Kartini reactor has been evaluated based on instrumentation system aspect. The Kartini reactor is designed by BATAN. Design power of the reactor is 250 kW, but it is currently operated at 100 kW. Instrumentation and control system function is to monitor and control the reactor operation. Instrumentation and control system consists of safety system, start-up and automatic power control, and process information system. The linear power channel and logarithmic power channel are used for measuring power. There are 3 types of control rod for controlling the power, i.e. safety rod, shim rod, and regulating rod. The trip and interlock system are used for safety. There are instrumentation equipment used for measuring radiation exposure, flow rate, temperature and conductivity of fluid The system of Kartini reactor has been developed by introducing a process information system, start-up system, and automatic power control. It is concluded that the instrumentation of Kartini reactor has followed the requirement and standard of IAEA. (author)

  3. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  4. Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area

    International Nuclear Information System (INIS)

    Yazid, M; Sutrisno; Sukarman-Aminjoyo; Zaenal-Abidin

    1996-01-01

    Evaluation of environmental radioactivity monitoring data around the Kartini Reactor area has been done. The aim of this investigation is for tracing the possibility of radioactivity released in the environment during the operation of Kartini reactor. The data was evaluated were monthly monitored data taken from 1986 to 1994 period. The method of analysis was done by comparing the environmental radioactivity data before and after reactor commissioning, off side the reactor up to a radius of 5.000 meters and more than 5.000 meters from Kartini reactor and also compared to the maximum permissible radioactivity according to the current regulation. This evaluation showed that there was no indication of radioactivity release to the environment during this period of reactor operation

  5. Evaluation of radiological impacts for environmental radioactivity distribution in the Kartini reactor area

    International Nuclear Information System (INIS)

    Yazid, M.; Suratman; Sutresna, G.; Aminjoyo, S.

    1996-01-01

    This evaluation covered of gross radioactivity. K-40 radioactivity in the water, soil and grass samples. The measurement of Cs-137 and Sr-90 radioactivity in the water samples have also been done. The aim of this research was determined of radiological impacts in the environment around the Kartini reactor. The water, soil and plant samples were counted of gross beta, Sr-90 activity by beta counter and for K-40, Cs-137 activity by low background gamma spectrometer. For this evaluation can be concluded there are no indication of the radioactivity release from the Kartini reactor operation. Gross beta radioactivity in the water, soil and grass sample are between 0.06-0.61 Bq/l, 0.24-0.79 Bq/g and 3.47-5.70 Bq/g ash. Radioactivity of K-40 in the water, soil and grass sample are between 0.09-0.56 Bq/l, 0.12-0.59 Bq/g and 0.29-0.93 Bq/g ash. The radioactivity of Cs-137 in the water samples are between under limit detectable level to 88.62 mBq/l and Sr-90 are between under limit detectable level to 24.22 mBq/l. (author)

  6. Design of Kartini reactor radiation monitor system using lab view

    International Nuclear Information System (INIS)

    Adi Abimanyu; Jumari; Achmad Fahrul Aji; Muhammad Khoiri

    2014-01-01

    Kartini Reactor operation will result in radiation exposure. Gamma radiation exposure rate at the Kartini Reactor monitored by several radiation monitors (Ludlum) that integrate with the computer, so that the rate of radiation exposure is always monitored. Current monitoring system combines six radiation monitor in one computer monitor radiation, and monitoring performed by operators and supervisors to see how the radiation exposure rate measured in the area around the reactor core in a periodic time manually. This research will develop a system to monitor radiation exposure in Kartini reactor based ATMega8 micro controller for interface between radiation monitor and computer and also Graphical User Interface (GUI) develop using Lab view software that makes monitoring is easier and documented regularly. This system is testing by simulation, it is done by replacing the function of the radiation monitoring devices (Ludlum) in Kartini Reactor with computers that send serial data with the same format with a format that is sent by Ludlum. The results show that the interface system has the ability to operate in a range of baud rate 1,200 bps, 2,400 bps, 4,800 bps, 9,600 bps, 14,400 bps, 19,200 bps and 38,400 bps, with the ability to provide realtime information every 6 seconds and able to document the rate of exposure to radiation in the form of logbook. (author)

  7. The Expert System For Safety Assesment Of Kartini Reactor Operation And Maintenance

    International Nuclear Information System (INIS)

    Syarip

    2000-01-01

    An expert system for safety assessment of Kartini reactor operation and maintenance based on fuzzy logic method has been made. The expert system is developed from the Fuzzy Expert System Tools (FEST), i.e. by developing the knowledge base and data base files of Kartini research reactor system and operations with an inference engine based on FEST. The knowledge base is represented in the procedural knowledge as heuristic rules or generally known as rule-base in the from of If-then rule. The fuzzy inference process and the conclusion of the rule is done by FEST based on direct chaining method with interactive as well as non-interactive modes. The safety assessment of Kartini reactor based on this method gives more realistic value than the conventional method or binary logic

  8. 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)

  9. REPRESENTASI NILAI-NILAI AL QURAN DALAM NOVEL KARTINI KARYA ABIDAH EL KHALIEQY

    Directory of Open Access Journals (Sweden)

    Haniah Haniah

    2018-04-01

    Full Text Available Penelitian ini bertujuan mengkaji nilai-nilai al quran yang ada di dalam novel Kartini karya Abidah El Khalieqy. Pengkajian ini didasari karena terdegradasinya moral generasi saat ini. Penurunan kualitas moral tersebut dilatarbelakangi oleh lemahnya keimanan di dalam diri seiring arus globalisasi yang tidak terbendung lagi. Penelitian ini adalah penelitian kualitatif. Peneliti akan mendeskripsikan hasil penelitian dengan rinci. Sumber data dari penelitian ini adalah novel Kartini karya Abidah El Khalieqy, sedangkan datanya adalah kutipan-kutipan teks di dalam novel yang mencerinkan nilai-nilai al quran.  Teknik analisis data yang digunakan adalah teknik analisis interaktif yang dikembangkan oleh Miles dan Huberman. Hasil penelitian ini menunjukan bahwa nilai-nilai al quran yang tercermin di dalam novel Kartini di antaranya kewajiban membaca, mencintai al quran, berbakti kepada orang tua, berserah diri kepada Allah, beribadah sebaik-baiknya, dan keutamaan menjadi orang yang berilmu.

  10. Design Engineering of Safety Parameter Display for Kartini Reactor

    International Nuclear Information System (INIS)

    Yoyok Dwi Setyo Pambudi; Suharyo Widagdo; Aliq Zuhdi; Darlis

    2003-01-01

    Modification information parameter of Reactor Kartini has been conducted. This program use compiler LabView 5.0 that compatible and relevant with National Instrument card. These card mostly use in reactor instrumentation and industry. The information that been display is control rod position, fuel temperature, and water pH. The development in this research is control rod figure that can move up and down as the output from reactor power. Beside this is a module-module security figure that can light if there is a warning signal. Maximum high of control rod is 100 cm. (author)

  11. Safety culture and quality management of Kartini research reactor

    International Nuclear Information System (INIS)

    Syarip; Hauptmanns, Ulrich

    1999-01-01

    The evaluation for assessing the safety culture and quality of safety management of Kartini research reactor is presented. The method is based on the concept of management control of safety (audit) as well as by using the developed method i.e. the questionnaires concerning areas of relevance which have to be answered with value statements. There are seven statements or qualifiers in answering the questions. Since such statements are vague, they are represented by fuzzy numbers. The weaknesses can be identified from the different areas contemplated. The evaluation result show that the quality of safety management of Kartini research reactor is globally rated as 'Average'. The operator behavior in the implementation of 'safety culture' concept is found as a weakness, therefore this area should be improved. (author)

  12. Measurement and evaluation of the external radiation level at reactor Kartini

    International Nuclear Information System (INIS)

    Atok Suhartanto; Suparno

    2013-01-01

    Measurement and evaluation of external radiation level at reactor Kartini in 2012 has been done. The purpose of this activity is to know the external radiation level as a result of the radioactive or radiation source usage, toward the operational of limit condition. The measurement is using survey meter Inspector 11086, factor of calibration 0.991 mR/h, at 9 locations is: Control room area, Thermal column facilities, Demineralizer, Beamport radiography facilities, bulk shielding Deck, Subcritical facilities, Reactor hall, Deck reactor and on the surface of reactor water tank . The highest room average measurement result in 9 working areas for 12 months continuously are at the reactor tank location is between 13.05±1.09 (xlO -2 mSv/hour) to 16.80±1.40 (x10 -2 mSv/hour), and the lowest measurement result in 1 location (control room) is 0.02±0.005 (x10 -2 mSv/hour) to 0.035±0.009 (x10 -2 mSv/hour). The Kartini reactor is involved in the control area which has potentially contaminated and has radiation exposure at the level of 6 mSv/year. Radiation Protection Officer that work in interval will received radiation exposure dosage of 8.4 mSv/year. This dosage is still below the Below Dosage Value which is recommended by, BAPETEN decree No, 4, 2013 about Protection and Radiation Safety in Nuclear Energy Application at 20 mSv/year. The result of the evaluation above shows that the external radiation which occurred in each area is still below the operational of limit condition that is written on the Kartini reactor safety analysis report, on document number: C7/05/B2/LAK/2010, revision 7. So that the workplace is safe for work monitored. (author)

  13. Safety culture and quality management of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia); Hauptmanns, Ulrich [Department of Plant Design and Safety, Otto-Von-Guericke-University, Magdeburg (Germany)

    1999-10-01

    The evaluation for assessing the safety culture and quality of safety management of Kartini research reactor is presented. The method is based on the concept of management control of safety (audit) as well as by using the developed method i.e. the questionnaires concerning areas of relevance which have to be answered with value statements. There are seven statements or qualifiers in answering the questions. Since such statements are vague, they are represented by fuzzy numbers. The weaknesses can be identified from the different areas contemplated. The evaluation result show that the quality of safety management of Kartini research reactor is globally rated as 'Average'. The operator behavior in the implementation of 'safety culture' concept is found as a weakness, therefore this area should be improved. (author)

  14. Monitoring of gross beta radioactivities on water sample environment in the surrounding of kartini reactor at 2011

    International Nuclear Information System (INIS)

    Siswanti; Munandar, A. Aris

    2013-01-01

    Measurement of gross beta radioactivities on water environment were done in the PTAPB BATAN has a goal for routine monitoring, with the result that fill RPL has been made and the result equivalented with quality standard were decided by BAPETEN. The water sample taken as much as 2 liter at 18 area were definited on radius 100 m to 5000 m in the surrounding of kartini reactor, vaporin on electric stove till the volume been ± 10 ml, and than pick out to the aluminium planset and drying on hot plate. Sample in the plancet were counted with a Low Background Counter (LBC) for 30 minutes and accounted of gross beta radioactivity water system. The result of gross beta radioactivity water environment at 2011 has a lowest 009, ± 0,06 Bq/I on Tambak Bayan area at june and in the Janti area highest 0,39 ± 0,08 Bq/ at December. The result still under of quality standard were decided by SK BAPETEN. No. 02/Ka- BAPETEN/V-99 is 0,4 Bq/I. (author)

  15. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  16. 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)

  17. Preliminary studi on neutronic aspect of a conceptual design of the Kartini reactor base ADS facility

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

    A preliminary study on neutronic aspect of a conceptual design of ADS facility with the basis of Kartini Reaktor, has been performed. The study was intended to see the feasibility from neutronic point of view of Kartini reactor, to be used as a small scale of NPP’s waste transmutation experimental facility. A SRAC code was used as the basis of calculations. The results indicate that the presence of minor actinides (MA) will give a positive reactivity, which tends to increase with the increase of MA concentrations. Based on the defined criteria of subcriticality and by considering the core power distributions and the level of reactivity contribution of MA element, it is concluded that Kartini reactor is potential enough to be used as an ADS experimental facility, mainly for MA concentration between 30 to 50 % of the assumed mixture of C-MA matrix. (author)

  18. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

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

  19. Kartini Research Reactor prospective studies for neutron scattering application

    International Nuclear Information System (INIS)

    Widarto

    1999-01-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10 7 n/cm 2 s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10 9 n/cm 2 s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  20. Pengaruh Pemindahan Perpustakaan Umum Kabupaten Rembang ke Lokasi Pariwisata Pantai Kartini Terhadap Minat Kunjung Pemustaka

    Directory of Open Access Journals (Sweden)

    Athiyah Kamaliyah

    2016-03-01

    Full Text Available The aim ofthis research are (1 to know the influence of Rembang Public Library transfer to Kartini Beach tourism area toward animo of visits to the libratyt (2 to know user perception about Rembang Public Library transfer to Kartini Beach tourism area (3 to know animo of visits to Rembang Public Library. This was a quantitative research that used correlation method between location (independent variable and animo of visits (dependent variable. Data were collected through questionnaire, interview and documentation. Data analysis used product moment correlation and processed using SPSS version 17 for Windows. The result of the study showed there was significant correlation between location and animo of visits. Score of correlation was 0.632 at significance level 5% r table is 0.202. The Rembang public library transfer to Kartini Beach tourism area belonged to good category whereby grand mean of location variable was 2.86. Animo of visits also belonged to high category whereby grand mean was 3.01. Key word: location, animo ofvisit, public library

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

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

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

  2. The Calculation Of Total Radioactivity Of Kartini Reactor Fuel Element

    International Nuclear Information System (INIS)

    Budisantoso, Edi Trijono; Sardjono, Y.

    1996-01-01

    The total radioactivity of Kartini reactor fuel element has been calculated by using ORIGEN2. In this case, the total radioactivity is the sum of alpha, beta, and gamma radioactivity from activation products nuclides, actinide nuclides and fission products nuclides in the fuel element. The calculation was based on irradiation history of fuel in the reactor core. The fuel element no 3203 has location history at D, E, and F core zone. The result is expressed in graphics form of total radioactivity and photon radiations as function of irradiation time and decay time. It can be concluded that the Kartini reactor fuel element in zone D, E, and F has total radioactivity range from 10 Curie to 3000 Curie. This range is for radioactivity after decaying for 84 days and that after reactor shut down. This radioactivity is happened in the fuel element for every reactor operation and decayed until the fuel burn up reach 39.31 MWh. The total radioactivity emitted photon at the power of 0.02 Watt until 10 Watt

  3. 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

  4. Design and Development of Data Acquisition System Process Parameters of Kartini Reactor

    International Nuclear Information System (INIS)

    Prajitno

    2009-01-01

    Design and development of computer program for data acquisition system of process parameters of the Kartini reactor have been done. System was designed using industrial computer which equipped with electronic module PCL-812PG. The function of computer is to take parameter data of reactor process, processing the data and displaying on the numeric form and bar graphic. Electronics module PCL- 12PG was installed in one of computer slot, functions to convert from analog signal to digital, received digital status signal and produce digital output. The analog signal and digital status got from logarithmic power channel, linear power channel dan three control rod. Result of data acquisition is merged in the form of ASCII characters block, send to the master computer serially with communications protocols RS-232. Computer program which has been developed was tested and used for monitoring Kartini reactor operation and give good performance result. (author)

  5. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

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

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

    International Nuclear Information System (INIS)

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

    1989-12-01

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

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

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

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

  8. 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.)

  9. 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.

  10. Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

    Science.gov (United States)

    Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

    2018-02-01

    Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

  11. 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)

  12. 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)

  13. 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

  14. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

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

  15. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  16. Safety Management and Safety Culture Self Assessment of Kartini Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip, S., E-mail: syarip@batan.go.id [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

    2014-10-15

    The self-assessment of safety culture and safety management status of Kartini research reactor is a step to foster safety culture and management by identifying good practices and areas for improvement, and also to improve reactor safety in a whole. The method used in this assessment is based on questionnaires provided by the Forum for Nuclear Cooperation in Asia (FNCA), then reviewed by experts. Based on the assessment and evaluation results, it can be concluded that there were several good practices in maintaining the safety status of Kartini reactor such as: reactor operators and radiation protection workers were aware and knowledgeable of the safety standards and policies that apply to their operation, readily accept constructive criticism from their management and from the inspectors of regulatory body that address safety performance. As a proof, for the last four years the number of inspection/audit findings from Regulatory Body (BAPETEN) tended to decrease while the reactor utilization and its operating hour increased. On the other hands there were also some comments and recommendations for improvement of reactor safety culture, such as that there should be more frequent open dialogues between employees and managers, to grow and attain a mutual support to achieve safety goals. (author)

  17. Analysis of Radioactivity Contamination Level of Kartini Reactor Efluen Gas to the Environment

    International Nuclear Information System (INIS)

    Suratman; Purwanto; Aminjoyo, S

    1996-01-01

    The analysis of radioactivity contamination level of Kartini reactor efluen gas to the environment has been done from 13-10-'95 until 8-2-'96. The aim of this research is to determine the radioactivity contamination level on the environment resulted from the release of Kartini reactor efluen gas and other facilities at Yogyakarta Nuclear Research Centre through stack. The analysis methods is the student t-test, the first count factor test and the gamma spectrometry. The gas sampling were carried out in the stack reactor, reactor room, environment and in other room for comparison. Efluen gas was sucked through a filter by a high volume vacuum pump. The filter was counted for beta, gamma and alpha activities. The radioactivity contamination level of the efluen gas passing through the stack to the environment was measured between 0.57 - 1.34 Bq/m3, which was equal to the airborne radioactivity in environment between 0.69 - 1.12 Bq/m3. This radioactivity comes from radon daughter, decay products result from the natural uranium and thorium series of the materials of the building

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

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

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

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

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

  1. 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)

  2. 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)

  3. Kartini's Children: On the Need for Thinking Gender and Education Together on a World Scale

    Science.gov (United States)

    Connell, Raewyn

    2010-01-01

    A world policy agenda for gender equality in education now exists, realising the idea of earlier reformers such as Kartini. This agenda, however, makes assumptions that are strongly contested by research and policy debates in national forums. This essay urges shifting the framework of gender analysis to global scale. It outlines what is involved…

  4. 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

  5. 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)

  6. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

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

  7. 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.

  8. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

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

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

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

  13. Evaluation of Radiological Impacts on the Operating Kartini Reactor and Natural Radioactivity of the Site Plan of Nuclear Power Plant Area

    International Nuclear Information System (INIS)

    Yazid, M; Sutresna, G; Sulistyono, A; Ngasifudin

    1996-01-01

    This radiological impacts evaluation covered of radioactivity in water, soil, grass, air samples and ambient gamma radiation that have been carried out in the Kartini reactor area and in the site plan of nuclear power plan are at Ujung Lemah Abang, Jepara, Central Java. The aim of this research was to determine that radiological impacts in the environment around the Kartini reactor compared to natural radioactivity for site plan of nuclear power plan area. The radioactivity in the water, soil and grass samples ware measured by low background beta counting system and were identified by low background gamma spectrometer. The radioactivity in the air samples was measured by beta portable counting system and the ambient gamma radiation was measured by portable high pressurized ionization chamber model RSS-112 Reuther-Stokes. The reactor data measurement was compared to the site plan of nuclear power plant area data for evaluation of radiological impacts on the operating reactor. From the evaluation and comparison can be concluded there are no indication of the radionuclide release from the reactor operation. The average radiactivity in the water, soil grass and air sample from the reactor area were between 0.17 - 0.61 Bq/1; 0,47 - 0,74 Bq/g; 4.43 - 4.60 Bq/g.ash and 49.53 - 70.90 x 10 Bq/cc. The average radioactivity of those sample from the nuclear power plant area were between 0.06-0.90 Bq/I; 0.02-0.86 Bq/g; 1.68-8.07 Bq/g.ash and 65.0-152.3 x 10 Bq/cc. The ambient gamma radiation were between 6.9-36.7 urad/h for the reactor area and 6.8-19.2 urad/h for the nuclear power plant area

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Science.gov (United States)

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

    2017-10-01

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

  16. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-12-01

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

  18. 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

  19. 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.)

  20. 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.)

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

    International Nuclear Information System (INIS)

    Felicione, F. S.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

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

  3. 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.)

  4. 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)

  5. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

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

  6. Ageing Management Programme: An Experience of In-Service Inspection of the Kartini Heat Exchanger

    Energy Technology Data Exchange (ETDEWEB)

    Nitiswati, S., E-mail: nitis@batan.go.id [Centre for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (BATAN), Jakarta (Indonesia); Syarip,; Tjiptono, T.; Wantana, [Centre for Accelerator and Material Process Technology, National Nuclear Energy Agency (BATAN), Yogyakarta (Indonesia)

    2014-08-15

    This paper discusses an experience on ISI of the Kartini reactor heat exchanger, as part of the implementation of an ageing management programme. Kartini reactor is located in Yogyakarta, Indonesia. The heat exchanger was constructed for 250 kW capacity. The type of heat exchanger is shell with tube recirculation. Tube material is stainless steel 304, and the shell and baffle plate materials are carbon steel. The heat exchanger has 72 tubes, its outer and inner diameters respectively are 19 and 16 mm, and tube thickness is 1.5 mm. The aim of ISI was to obtain and evaluate the heat exchanger’s condition including the detection of any possible local tube thinning, pitting corrosion or gradual thinning, and determine whether any degradation or deterioration of the heat exchanger could have a significant impact to safety. The heat exchanger was inspected utilizing eddy current equipment in 2003 and 2006. Inspection results in 2003 determined that 12 heat exchanger tubes have a thinning degradation level ranging from 10% up to 60% of the outer diameter due to pitting corrosion. Deterioration of baffle plates has been linked to general corrosion attack. Inspection results in 2006 showed a consistent thinning degradation level with the previous inspection in 2003. So far heat exchanger performance is still satisfactory, as defined by the transfer of primary heat for a 2°C difference between inlet and outlet as required. (author)

  7. 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

  8. Design of Computerized in Core Fuel Management System of Kartini Reactor

    International Nuclear Information System (INIS)

    Edi-Trijono-Budisantoso; Sardjono, Y; Edi-Purwanto; Widi-Setiawan

    2000-01-01

    The program organization for managing Kartini reactor fuel elements has been designed. This program organization work to process on-line operationdata-base and core configuration data-base to produce data-base for in-corefuer management. The in-core fuel management data-base consist of irradiationhistory card, radionuclides inventory and radiation dose for each fuelelement. The computation in this process based on the ORIGEN2, TRIGAP codesand some in-house developed codes that perform matching between output dataof many codes to output data of other code. This program organization worksunder control of a program manager by following the scheduled time table. Thedesign gives a description of the first step development of the in-core fuelmanagement that will be implemented in the internet web server. (author)

  9. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tjiptono, Tri Wulan; Syarip,

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  10. Kepuasan Kerja dan Dampaknya terhadap Unjuk Kerja Guru di Gugus Pantai Kartini UPT Dinas Pendidikan Kecamatan Mojolaban (Tinjauan Aspek Keteladanan Pimpinan dan Ketrampilan Pemanfaatan Media

    Directory of Open Access Journals (Sweden)

    Emi Sugiati

    2015-07-01

    Full Text Available This research aims to describe : 1 the contribution of leader exemplary and learning media model through indirect satisfied concerning the characteristic of the teacher performance, 2 The contribution of leader exemplary and the satisfied learning media model, 3 the satisfied of the teacher performance. The result is expected to gave information to the headmaster, so its can become reconciliation to increase the school work and for the teacher they can get a feet-back about the increase of the work. This quantitative research has used inquiry method that proved valid and reliable and documentations method a means of principal data collecting, and to test direct and indirect contribution between keteladanan leadership, study media, satisfaction and performance teacher at coast cluster Kartini UPT district Mojolaban education official use stripe analysis method (path Analysis. population in these watchfulness teachers coast cluster Pantai Kartini UPT district education official Mojolaban as much as 141 person, with sample as much as 106 person. The result show that : 1 The leader exemplary and learning media give direct contribution for the teacher performance, 2 The leader exemplary and learning media give direct significant to the satisfied, 3 The satisfied give direct contribution and significant for the performance of the teacher in coast cluster Pantai Kartini Education Official UPTD Mojolaban sub district.

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

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Jacopo Buongiorno

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    2008-12-01

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

  15. 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.)

  16. 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.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

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

    Science.gov (United States)

    Zapata, Luis E.; Hackel, Lloyd

    1999-01-01

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

  19. 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.

  20. 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)

  1. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Science.gov (United States)

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

    2014-01-01

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

  3. 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.

  4. 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

  5. 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

  6. Finite Element Method in the Three Dimensions Deformation Computation ofKartini Reactor Stack

    International Nuclear Information System (INIS)

    Supriyono; Syarip; Wibisono, I

    2000-01-01

    The calculation of the Kartini reactor stack i.e. one of the nuclearinstallations in P3TM-BATAN Yogyakarta by using SAP 90 software have beendone. The calculation is done as a safety review of building towards theearthquake style in Yogyakarta. The 3-dimension deformation calculation isperformed by the numeric method i.e. finite element method with the form ofelements is the shell. The result obtained showed that the construction oftower safe to the existing earthquake, where the moment exerted as a resultof earthquake style was different under the moment having been kept by thebuilding structure. By knowing the deformation on the stack it is expectedcould be used for concluding the strength of the whole reactor building.(author)

  7. 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)

  8. 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

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

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1994-01-01

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

  10. Determination and Analysis of Ar-41 Dose Rate Characteristic at Thermal Column of Kartini Reactor

    International Nuclear Information System (INIS)

    Widarto; Sardjono, Y.

    2007-01-01

    Determination and Analysis of Ar-41 activity dose rate at the thermal column after shutdown of Kartini reactor has been done. Based on evaluation and analysis concluded that external dose rate is D = 1.606x10 -6 Sv/second and internal dose rate is 3.429x10 -1 1 Sv/second. It means that if employee work at the column thermal area for 15 minutes a day, 5 days a week, in a year will be 0.376 Sv still under dose rate limit i.e. 0.5 Sv, so that the column thermal facility is safely area. (author)

  11. 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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  13. 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.)

  14. 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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  16. 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)

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

    International Nuclear Information System (INIS)

    Gu Yingbin

    2012-01-01

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

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

    Science.gov (United States)

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

    2003-06-01

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

  19. 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)

  20. 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.)

  1. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-02-01

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

  4. 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

  5. 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

  6. 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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  8. 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%

  9. 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.

  10. 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

  11. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  12. 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

  13. 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

  14. Application Of WIMS Code To Calculation Kartini Reactor Parameters By Pin-Cell And Cluster Method

    International Nuclear Information System (INIS)

    Sumarsono, Bambang; Tjiptono, T.W.

    1996-01-01

    Analysis UZrH fuel element parameters calculation in Kartini Reactor by WIMS Code has been done. The analysis is done by pin cell and cluster method. The pin cell method is done as a function percent burn-up and by 8 group 3 region analysis and cluster method by 8 group 12 region analysis. From analysis and calculation resulted K ∼ = 1.3687 by pin cell method and K ∼ = 1.3162 by cluster method and so deviation is 3.83%. By pin cell analysis as a function percent burn-up at the percent burn-up greater than 59.50%, the multiplication factor is less than one (k ∼ < 1) it is mean that the fuel element reactivity is negative

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

    Science.gov (United States)

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

    2017-08-01

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

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

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

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

  17. 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

  18. 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.

  19. 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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  1. 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)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. 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

  4. Gamma Radiation Assessment In Kartini Reactor And Its Vicinity

    International Nuclear Information System (INIS)

    Yazid, M.; Supriyatni, E.; Maryono; Bastianudin, Aris

    2000-01-01

    Measurement to calculate dose assesment for gamma radiation in Kartini Reactor and its vicinity has been done whether on operated or un operated condition. Measurement was performed using height pressured ionization chamber, Reuther Stokes RS-112 production. Measurement location was determined based on distance variation inwardly and outwardly of reactor building and its vicinity. The result showed that the average dose rate in the reactor building when un operated is in the range of 11.4-38.6 mu rad/hour and when the reactor operated is 166.4-1910.9 mu rad/hour. While the vicinity of the reactor on operated condition the average dose rate is 34.4-38.6 mu rad/hour in un operated condition is 6.9-7.0 mu rad/hour. This result showed that the reactor operated did not rise the radiation exposure level in its vicinity. From the personnel assesment dose rate of gamma radiation is 28.54 mrem/week on operated condition, 0.90 mrem.week on un operated condition. While dose rate outside the reactor is 0.44 and 0.27 mrem/week for operated and un operated condition consecutively. This dose rate is still below maximum permissible dose than recommended by the national regulation of radiation protection from BAPETEN No. 01/Ka.BAPETEN/V-99

  5. 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

  6. 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.)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Dattatraya G. Subhedar

    2018-03-01

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

  10. 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

  11. 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

  12. 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

  13. 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.)

  14. 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.)

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

    International Nuclear Information System (INIS)

    Santos, G.A. dos.

    1990-01-01

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

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

    Science.gov (United States)

    Freche, John C; Schum, Eugene F

    1951-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

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

  20. 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

  1. 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.)

  2. 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.)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  4. 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

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

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

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

  6. 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)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    Science.gov (United States)

    Diaguila, Anthony J; Freche, John C

    1951-01-01

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

  10. 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.)

  11. 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

  12. 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.)

  13. 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.)

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

    Directory of Open Access Journals (Sweden)

    Senthilraja S.

    2017-03-01

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

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

    International Nuclear Information System (INIS)

    Johnson, K.R.

    1981-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  17. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  19. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  1. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

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

  2. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

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

  3. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

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

  4. 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.

  5. 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.

  6. 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

  7. 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.

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

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  10. Contact condensation effects in the main coolant pipe

    International Nuclear Information System (INIS)

    Haefner, W.; Fischer, K.

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    2018-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    MARSHALL, THERON D.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo

    1979-03-01

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

  17. 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.)

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  19. Kartini reactor tank inspection using NDT method for safety improvement of the reactor operation

    International Nuclear Information System (INIS)

    Syarip; Sutondo, Tegas; Saleh, Chaerul; Nitiswati; Puradwi; Andryansah; Mudiharjo

    2002-01-01

    The inspection of Kartini reactor tank liner (TRK) by using Non Destructive Testing (NDT) methods to improve the reactor operation safety, have been done. The type of NDT used were: visual examination using an underwater camera and magnifier, replication survey using dental putty, hardness test using an Equotip D indentor, thickness test using ultrasonic probe, and dye penetrant test. The visual examination showed that the surface of TRK was in good condition. The hardness readings were considered to be consistent with the original condition of the tank and the slight hardness increase at the reactor core area consistent with the neutron fluence experienced -10 1 4 n/cm 2 . Results of ultrasonic thickness survey showed that in average the TRK thickness is between 5,0 mm - 6,5 mm, a low 2,1 mm thickness exists at the top of the TRK in the belt area (double layer aluminum plat, therefore do not influencing the safety ). The replica and dye penetrant test at the low thickness area and several suspected areas showed that it could be some defect from original manufacture. Therefore, it can be concluded that the TRK is still feasible for continued operation safely

  20. 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

  1. 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

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

    International Nuclear Information System (INIS)

    Elain, L.

    2004-12-01

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

  3. 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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Corradini, M.L.

    1983-01-01

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

  6. The effect of leadership, organizational culture, and competency on teachers' performance in Ibu Kartini vocational high school Semarang

    Science.gov (United States)

    Toha, Mohamad; Katoningsih, Sri

    2018-03-01

    The low performance of teachers in organization is influenced many factors. Organizational culture could be the key of organization success; hence many researches were done to identify the value and the attitude norm that gave big contribution for organization success. Competency is a part of employee they perform during work as kind of behavior. Competency depends on the aspects process of teachers' performance. The purpose of this research is to know the effect of leadership, organizational culture and competency on teachers' performance. The objects of this research are leadership, organizational culture, competency and teachers' performance in Ibu Kartini vocational high school. This research is quantitative. To collect the data, questionnaire was used. Then, the data were analyzed by using Path analysis in SPPS 16. The result of this research showed that leadership, organizational culture, competency and performance run well and had significant effect on teachers' performance.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Tsutsui, Toshiaki

    1979-01-01

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

  9. Coolant radiolysis studies in the high temperature, fuelled U-2 loop in the NRU reactor

    International Nuclear Information System (INIS)

    Elliot, A.J.; Stuart, C.R.

    2008-06-01

    An understanding of the radiolysis-induced chemistry in the coolant water of nuclear reactors is an important key to the understanding of materials integrity issues in reactor coolant systems. Significant materials and chemistry issues have emerged in Pressurized Water Reactors (PWR), Boiling Water Reactors (BWR) and CANDU reactors that have required a detailed understanding of the radiation chemistry of the coolant. For each reactor type, specific computer radiolysis models have been developed to gain insight into radiolysis processes and to make chemistry control adjustments to address the particular issue. In this respect, modelling the radiolysis chemistry has been successful enough to allow progress to be made. This report contains a description of the water radiolysis tests performed in the U-2 loop, NRU reactor in 1995, which measured the CHC under different physical conditions of the loop such as temperature, reactor power and steam quality. (author)

  10. 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

  11. Lubrication analysis of the journal bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Kim, J. I.; Jang, M. H.

    2000-01-01

    Special type journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel. The MCP bearings are lubricated with water without external lubricating oil supply. Long bearing with vertical grooves is designed with relatively large bearing clearance to accommodate the long shaft. Lubricational analysis method for journal bearing with vertical grooves in the main coolant pump of SMART is proposed, and lubricational characteristics of the bearings are examined in this paper

  12. 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

  13. 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

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  15. Method of suppressing the deposition of Co-60 to primary coolant pipeways in a nuclear reactor

    International Nuclear Information System (INIS)

    Hoshi, Michio; Tachikawa, Enzo; Goto, Satoshi; Sagawa, Chiaki; Yonezawa, Chushiro.

    1987-01-01

    Purpose: To suppress the deposition of Co-60 to primary coolant pipeways in a nuclear reactor. Method: To reduce the accumulation of Co-60 by causing chemical species of extremely similar chemical property with soluble Co-60 to be present together in coolants and replacing the deposition of Co-60 to the primary coolant pipeways in a nuclear reactor with that of the coexistent chemical spacies. Ni or Zn is used as the coexistet chemical spacies of similar chemical property with Co-60. The coexistent amount is from 5 to 10 times of the soluble Co-60 in the primary coolants. Ni or Zn solution adjusted with concentration is poured into and mixed with the coolants from a water feed source by using a high pressure constant volume pump. The amount of Co-60 taken into the pipeways caused by corrosion due to high temperature coolant is reduced to about 1/5 as compared with the case of Co-60 alone if 1 ppb of soluble Co-60 is present in water and 5 ppb of soluble Ni or Zn is added and, reduced to 1/12 if the amount of Ni or Zn is 10 ppb. (Kamimura, M.)

  16. 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

  17. Coolant clean-up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tsuburaya, Hirobumi; Akita, Minoru; Shiraishi, Tadashi; Kinoshita, Shoichiro; Okura, Minoru; Tsuji, Akio.

    1987-01-01

    Purpose: To ensure a sufficient urging pressure at the inlet of a coolant clean-up system pump in a nuclear reactor and eliminate radioactive contaminations to the pump. Constitution: Coolant clean-up system (CUW) pump in a nuclear reactor is disposed to the downstream of a filtration desalter and, for compensating the insufficiency of the urging pressure at the pump inlet, the reactor water intake port to the clean-up system is disposed to the downstream of the after-heat removing pump and the heat exchanger. By compensating the net positive suction head (NPSH) of the clean-up system from the residual heat removing system, the problems of insufficient NPSH for the CUW pump upon reactor shut-down can be dissolved and, accordingly, the reactor clean-up system can be arranged in the order of the heat exchanger, clean-up device and pump. Thus, the CUW pump acts on reactor water after cleaned-up in the clean-up device to reduce the radioactivity contamination to the pump. (Kawakami, Y.)

  18. Correlation of cylinder-head temperatures and coolant heat rejections of a multicylinder, liquid-cooled engine of 1710-cubic-inch displacement

    Science.gov (United States)

    Lundin, Bruce T; Povolny, John H; Chelko, Louis J

    1949-01-01

    Data obtained from an extensive investigation of the cooling characteristics of four multicylinder, liquid-cooled engines have been analyzed and a correlation of both the cylinder-head temperatures and the coolant heat rejections with the primary engine and coolant variables was obtained. The method of correlation was previously developed by the NACA from an analysis of the cooling processes involved in a liquid-cooled-engine cylinder and is based on the theory of nonboiling, forced-convection heat transfer. The data correlated included engine power outputs from 275 to 1860 brake horsepower; coolant flows from 50 to 320 gallons per minute; coolants varying in composition from 100 percent water to 97 percent ethylene glycol and 3 percent water; and ranges of engine speed, manifold pressure, carburetor-air temperature, fuel-air ratio, exhaust-gas pressure, ignition timing, and coolant temperature. The effect on engine cooling of scale formation on the coolant passages of the engine and of boiling of the coolant under various operating conditions is also discussed.

  19. Reduction of corrosion products in water coolant - basic way of increase in efficiency and improvement of ecological safety of NPU

    International Nuclear Information System (INIS)

    Prozorov, V.V.

    2004-01-01

    Corrosion of oxidated steel in water with additives of inhibitors or oxygen was considered. It is shown that preliminary oxidation of steel makes possible declining concentration of inhibitors or oxygen. Experiments demonstrate possibilities of the neutral-oxygen water regime for supply of the effective protection. Corrosion resistance of steel may be increased in many times through correct aqua-chemical regimes. Also concentration of corrosion products may be decreased in many times in coolant and their activation in neutron flux of nuclear reactor, amount of radioisotopes [ru

  20. 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

  1. 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

  2. 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.)

  3. Determination and analysis of neutron flux distribution on radial Piercing beam port for utilization of Kartini research reactor

    International Nuclear Information System (INIS)

    Widarto

    2002-01-01

    Determination and analysis of neutron flux measurements on radial piercing beam port have been done as completion experimental data document and progressing on utilization of the Kartini research reactor purposes. The analysis and determination of the neutron flux have been carried out by using Au foils detector neutron activation analysis method which put on the radius of cross section (19 cm) and a long of radial piercing beam port (310 cm) Based on the calculation, distribution of the thermal neutron flux is around (8.3 ± 0.9) x 10 5 ncm -2 s -1 to (6.8 ± 0.5) x 10 7 ncm -2 s -1 and fast neutron is (5.0 ± 0.2) x 10 5 ncm -2 s -1 to (1.43 ± 0.6) x 10 7 ncm -2 s -1 . Analyzing by means of curve fitting method could be concluded that the neutron flux distribution on radial piercing beam port has profiled as a polynomial curve. (author)

  4. 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)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  6. 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.)

  7. 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.

  8. Investigating Liquid CO2 as a Coolant for a MTSA Heat Exchanger Design

    Science.gov (United States)

    Paul, Heather L.; Padilla, Sebastian; Powers, Aaron; Iacomini, Christie

    2009-01-01

    Metabolic heat regenerated Temperature Swing Adsorption (MTSA) technology is being developed for thermal and carbon dioxide (CO 2) control for a future Portable Life Support System (PLSS), as well as water recycling. CO 2 removal and rejection is accomplished by driving a sorbent through a temperature swing of approximately 210 K to 280 K . The sorbent is cooled to these sub-freezing temperatures by a Sublimating Heat Exchanger (SHX) with liquid coolant expanded to sublimation temperatures. Water is the baseline coolant available on the moon, and if used, provides a competitive solution to the current baseline PLSS schematic. Liquid CO2 (LCO2) is another non-cryogenic coolant readily available from Martian resources which can be produced and stored using relatively low power and minimal infrastructure. LCO 2 expands from high pressure liquid (5800 kPa) to Mars ambient (0.8 kPa) to produce a gas / solid mixture at temperatures as low as 156 K. Analysis and experimental work are presented to investigate factors that drive the design of a heat exchanger to effectively use this sink. Emphasis is given to enabling efficient use of the CO 2 cooling potential and mitigation of heat exchanger clogging due to solid formation. Minimizing mass and size as well as coolant delivery are also considered. The analysis and experimental work is specifically performed in an MTSA-like application to enable higher fidelity modeling for future optimization of a SHX design. In doing so, the work also demonstrates principles and concepts so that the design can be further optimized later in integrated applications (including Lunar application where water might be a choice of coolant).

  9. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

  10. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  11. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  12. 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.)

  13. 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

  14. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  15. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-15

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

  17. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  18. Study of radiation exposure rate on the measurement points in Kartini reactor hall as based to determine operation safety parameters (KBO)

    International Nuclear Information System (INIS)

    Mahrus Salam; Elisabeth Supriyatni; Fajar Panuntun

    2016-01-01

    In the operation of nuclear facility there are safety parameters, which is the value of the conservatively maximum limit to ensure that all of the uncertainty in the analysis of facility operations safety have been considered, such as uncertainty of measurement, response time and uncertainty calculation tool, and is get a long to others value of normal operating condition limits, in other words, there are still allowed or permitted. Calculation of the radiation exposure rate on five measurement points (50 cm above the water surface of reactor pool, above interim storage (bulk shielding), reactor deck, thermal column and sub critical facility) and to be compared to the operation safety parameters (KBO) of Kartini reactor. The exposure rate value is obtained by calculating the source term of radioactivity on the core, attenuation resulting from the radiation shielding and measurement distance. From the calculation obtained that the value of gamma exposure rate of 50 cm above the water surface of reactor pool is 96.91 mR/hr (KBO<100 mR/hr), on the deck of Bulk Shielding amounted to 1.70 mR/h (KBO<2.5 mR/hr), on the reactor deck amounted to 5.73 mR/hr (KBO<10 mR/hr), on the Thermal Column amounted to 2.73 mR/hr (KBO<10 mR/hr) and on the sub critical facility amounted to 1.148 mR/hr (KBO<2.5 mR/hr). The value of gamma exposure rate at 5 locations measurements are still less than the operation safety parameters (KBO), it means that the reactor is safe to be operated. (author)

  19. 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

  20. 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

  1. Effects of molten material temperatures and coolant temperatures on vapor explosion

    Institute of Scientific and Technical Information of China (English)

    LI Tianshu; YANG Yanhua; YUAN Minghao; HU Zhihua

    2007-01-01

    An observable experiment facility for low-temperature molten materials to be dropped into water was set up in this study to investigate the mechanism of the vapor explosion. The effect of the fuel and coolant interaction(FCI) on the vapor explosion during the severe accidents of a fission nuclear reactor has been studied. The experiment results showed that the molten material temperature has an important effect on the vapor explosion behavior and pressure. The increase of the coolant temperature would decrease the pressure of the vapor explosion.

  2. 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

  3. 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.)

  4. Premixing and steam explosion phenomena in the tests with stratified melt-coolant configuration and binary oxidic melt simulant materials

    Energy Technology Data Exchange (ETDEWEB)

    Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se; Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Konovalenko, Alexander, E-mail: kono@kth.se; Karbojian, Aram, E-mail: karbojan@kth.se

    2017-04-01

    Highlights: • Steam explosion in stratified melt-coolant configuration is studied experimentally. • Different binary oxidic melt simulant materials were used. • Five spontaneous steam explosions were observed. • Instability of melt-coolant interface and formation of premixing layer was observed. • Explosion strength is influenced by melt superheat and water subcooling. - Abstract: Steam explosion phenomena in stratified melt-coolant configuration are considered in this paper. Liquid corium layer covered by water on top can be formed in severe accident scenarios with (i) vessel failure and release of corium melt into a relatively shallow water pool; (ii) with top flooding of corium melt layer. In previous assessments of potential energetics in stratified melt-coolant configuration, it was assumed that melt and coolant are separated by a stable vapor film and there is no premixing prior to the shock wave propagation. This assumption was instrumental for concluding that the amount of energy that can be released in such configuration is not of safety importance. However, several recent experiments carried out in Pouring and Under-water Liquid Melt Spreading (PULiMS) facility with up to 78 kg of binary oxidic corium simulants mixtures have resulted in spontaneous explosions with relatively high conversion ratios (order of one percent). The instability of the melt-coolant interface, melt splashes and formation of premixing layer were observed in the tests. In this work, we present results of experiments carried out more recently in steam explosion in stratified melt-coolant configuration (SES) facility in order to shed some light on the premixing phenomena and assess the influence of the test conditions on the steam explosion energetics.

  5. Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, B.R.

    1986-01-01

    In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

  6. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  7. Application of heat-resistant non invasive acoustic transducers for coolant control in the NPP pipelines

    International Nuclear Information System (INIS)

    Melnikov, V.; Nigmatulin, B.

    1997-01-01

    The use of ultrasonic waves enables remote testing of the coolant flow, detection of solid and gaseous occlusions and measuring of the water velocity and level. Analysis of the acoustic noise makes it possible to detect coolant leaks and diagnose the state and operation of the rotating mechanisms and bearings. Results are given of the research in the development of highly reliable waveguide-type non-invasive acoustic transducers with a long service life. Examples are given of the use of transducers in various fields of nuclear technology: detection of gas in coolant, indication of the coolant level, control of pipe filling and drainage, measurement of liquid film velocity at the pipe inner surface. (M.D.)

  8. 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)

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

    International Nuclear Information System (INIS)

    Finn, P.A.

    1991-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Finn, P.A.

    1991-03-01

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

  11. 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

  12. 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

  13. Coolant mixing in pressurized water reactors. Pt. 1. Feasibility of closed analytical solutions and simulation of the mixing with CFX-4. Final report

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Rohde, U.

    2001-10-01

    The project was aimed at the analytical and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing velocity measurements by the LDA technique. Design and construction of the ROCOM facility including the measurement equipment were performed in a second part of the project. For the design of the facility, CFD calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. A theoretical 2D-model of the downcomer flow was developed based on the potential theory. The coolant inlet is represented by mass sources. Potential vortices were superposed to describe large scale recirculations. However, the method requires an a-priory knowledge of the location and intensity of the vorticity sources. Therefore, the main goal of the project was the numerical simulation of the coolant mixing of different PWRs. The temperature and boron concentration fields established by the coolant mixing during nominal and transient flow conditions in the pressure vessel of the PWR Konvoi and the Russian type WWER-440 were investigated. The calculations were carried out with the CFD-code CFX 4. The results of the CFD calculation are found in the final report. The report is based on the Ph.D. work of T. Hoehne. (orig.) [de

  14. 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)

  15. 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

  16. Linear titration plot for the determination of boron in the primary coolant of a pressurized water reactor

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1992-01-01

    A linear titration plot method has been devised for the determination of boron as boric acid in partly neutralized solution, such as occurs in the primary coolant of pressurized water reactors. The total boron and the alkali in the sample are determined simultaneously. Although it is not essential to add mannitol in this method, it is more accurate when the solution is saturated with mannitol. Comparisons are made with other modes of titration: Gran plots, first and second differential potentiometric titrations and indicator titrations. None of these gives the total boron directly in partly neutralized solutions. (author)

  17. A open-quotes zero wasteclose quotes coolant management strategy

    International Nuclear Information System (INIS)

    Kennicott, M.A.

    1994-01-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant's Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant's liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study's goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a open-quotes zero wasteclose quotes machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs

  18. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  19. Lubrication analysis of the thrust bearing in the main coolant pump of SMART

    International Nuclear Information System (INIS)

    Lee, J. S.; Park, J. S.; Kim, J. H.; Hur, H.; Kim, J. I.

    2001-01-01

    Thrust bearing and journal bearings are installed in the main coolant pump for SMART to support the rotating shaft with proper lubrication. The canned motor type main coolant pumps are arranged vertically on the reactor vessel and especially the MCP bearings are lubricated with water without external lubricating oil supply. Because axial load capacity of the thrust bearing can hardly meet requirement to acquire hydrodynamic or fluid film lubrication state, self-lubrication characteristics of silicon graphite meterials would be needed. Lubricational analysis method for thrust bearing for the main coolant pump of SMART is proposed, and lubricational characteristics of the bearing generated by solving the Reynolds equation are examined in this paper

  20. Device for preventing leakage of coolant in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kobayashi, Yukio; Sekiguchi, Mamoru; Yoshida, Hideo.

    1975-01-01

    Object: To prevent leakage of coolant from between lower tie plate and channel box without causing deformation of the channel box and also without the possibility of disturbing the installation and removal of the box by the provision of a thin plate provided with leakage holes for the lower tie plate. Structure: Static water pressure within the lower tie plate is adapted to act upon the bear side of a flat plate for leakage prevention through leakage holes formed in the tie plate, thus urging the flat plate against the channel box inner surface. Meanwhile, static water pressure having been led through the leakage holes in the flat plate is adapted to press the flat plate in the vertical direction, thus urging the flat plate against the channel box inner surface and thereby preventing leakage of the coolant through a gap between the channel box and lower tie plate. (Yoshino, Y.)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  2. Some observations on simulated molten debris-coolant layer dynamics

    International Nuclear Information System (INIS)

    Greene, G.A.; Klein, J.; Klages, J.; Schwarz, E.; Sanborn, Y.

    1983-04-01

    Experiments are being performed to investigate high temperature liquid-liquid film boiling between a pool of liquid metal and an overlying coolant pool of R-11 or water. Film boiling has been observed to be stable for R-11; however, considerable liquid-liquid contact has been observed with water well beyond the minimum film boiling temperature. Unstable liquid-liquid film boiling of water has been observed to escalate into dispersive, non-energetic vapor explosions when the interface contact temperature exceeded the spontaneous nucleation temperature. Other parametric trends in the data are discussed

  3. Molten fuel/coolant interaction studies: some results obtained with the Windscale small shock tube rig

    International Nuclear Information System (INIS)

    Higham, E.J.; Vaughan, G.J.

    1978-02-01

    Experiments are described in which water has been brought into contact with various molten metals in a shock tube, thus simulating the fall of coolant into molten uranium dioxide in a postulated reactor accident. Impact velocities of the water on to the molten material were in the range 5 to 7 m/s. Shock-pulse pressures in the water column after impact and particle size distributions of the dispersed resolidified material that was recovered were measured. The proportion of dispersed material and the size of the shock pulse (by comparison with that expected from water hammer alone) have been used as criteria for the occurrence of a molten fuel/coolant interaction and such interactions of varying degrees of violence have been found for water/aluminium, water/bismuth, water/tin, over a range of temperatures from 350 0 C to 950 0 C, for water/boric oxide, but not for water/magnesium. (author)

  4. 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.

  5. Compatibility of structural materials with fusion reactor coolant and breeder fluids

    International Nuclear Information System (INIS)

    DeVan, J.H.

    1979-01-01

    Fusion reactors are characterized by a lithium-containing blanket, a heat transfer medium that is integral with the blanket and first wall, and a heat engine that couples to the heat transfer medium. A variety of lithium-containing substances have been identified as potential blanket materials, including molten lithium metal, molten LiF-BeF 2 , Pb-Li alloys, and solid ceramic compounds such as Li 2 O. Potential heat transfer media include liquid lithium, liquid sodium, molten nitrates, water, and helium. Each of these coolants and blankets requires a particular set of chemical and mechanical properties with respect to the associated reactor and heat engine structural materials. This paper discusses the materials factors that underlie the selection of workable combinations of blankets and coolants. It also addresses the materials compatibility problems generic to those blanket-coolant combinations currently being considered in reactor design studies. (orig.)

  6. Determination of primary flow by correlation of temperatures of the coolant

    International Nuclear Information System (INIS)

    Villanueva, Jose

    2003-01-01

    Correlation techniques are often used to assess primary coolant flow in nuclear reactors. Observable fluctuations of some physical or chemical coolant properties are suitable for this purpose. This work describes a development carried out at the National Atomic Energy Commission of Argentina (CNEA) to apply this technique to correlate temperature fluctuations. A laboratory test was performed. Two thermocouples were installed on a hydraulic loop. A stationary flow of water circulated by the mentioned loop, where a mechanical turbine type flowmeter was installed. Transit times given by the correlation flowmeter, for different flow values measured with the mechanical flowmeter, were registered and a calibration between them was done. A very good linear behavior was obtained in all the measured range. It was necessary to increase the fluctuation level by adding water at different temperatures at the measuring system input. (author)

  7. 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.)

  8. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  9. Deposition and incorporation of corrosion product to primary coolant suppressing method

    International Nuclear Information System (INIS)

    Tsuzuki, Yasuo; Hasegawa, Naoyoshi; Fujioka, Tsunaaki.

    1992-01-01

    In a PWR type nuclear power plant, the concentration of dissolved nitrogen in primary coolants is increased by controlling the nitrogen partial pressure in a volume controlling tank gas phase portion or addition of water in a primary system water supply tank containing dissolved nitrogen to a primary system. Then ammonium is formed by a reaction with hydrogen dissolved in the primary coolants in the field of radiation rays, to control the concentration of ammonium in the coolants within a range from 0.5 to 3.5 ppm, and operate the power plant. As a result, deposition and incorporation of corrosion products to the structural materials of the primary system equipments during plant operation (pH 6.8 to 8.0) are suppressed. In other words, deposition of particulate corrosion products on the surface of fuel cladding tubes and the inner surface of pipelines in the primary system main equipments is prevented and incorporation of ionic radioactive corrosion products to the oxide membranes on the inner surface of the pipelines of the primary system main equipments is suppressed, to greatly reduce the radiation dose rate of the primary system pipelines. Thus, operator's radiation exposure can be decreased upon shut down of the plant. (N.H.)

  10. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

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

    International Nuclear Information System (INIS)

    Puri, R.K.; Singh, M.

    2006-01-01

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

  12. 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

  13. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  14. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  15. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    International Nuclear Information System (INIS)

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

    1997-01-01

    This paper presents results of experimental studies on the heat transfer and solidifcation of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. As a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleight number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer

  16. Water systems

    International Nuclear Information System (INIS)

    Riess, R.

    1980-01-01

    The present paper describes the coolant chemistry and its consequences for 1300 MWsub(e) KWU PWR plants. Some selected systems, i.e. primary heat transport system, steam water cycle and cooling water arrangements, are chosen for this description. Various aspects of coolant chemistry regarding general corrosion, selective types of corrosion and deposits on heat transfer surfaces have been discussed. The water supply systems necessary to fulfill the requirements of the coolant chemistry are discussed as well. It has been concluded that a good operating performance can only be achieved when - beside other factors - the water chemistry has been given sufficient consideration. (orig./RW)

  17. Water systems

    International Nuclear Information System (INIS)

    Riess, R.

    1981-01-01

    The present paper describes the coolant chemistry and its consequences for 1300 MWsub(e) KWU PWR plants. Some selected systems, i.e. primary heat transport system, steam water cycle and cooling water arrangements, are chosen for this description. Various aspects of coolant chemistry regarding general corrosion, selective types of corrosion and deposits on heat transfer surface have been discussed. The water supply systems necessary to fulfill the requirements of the coolant chemistry are discussed as well. It has been concluded that a good operating performance can only be achieved when - beside other factors - the water chemistry has been given sufficient consideration. (orig./RW)

  18. In reactor performance of defected zircaloy-clad U{sub 3}Si fuel elements in pressurized and boiling water coolants

    Energy Technology Data Exchange (ETDEWEB)

    Feraday, M A; Allison, G M; Ambler, J F.R.; Chalder, G H; Lipsett, J J

    1968-05-15

    The results of two in-reactor defect tests of Zircaloy-clad U{sub 3}Si are reported. In the first test, a previously irradiated element ({approx}5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at {approx}270{sup o}C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U{sub 3}Si at 300{sup o}C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  19. Primary coolant feed and bleed operating regions for the Midland Plant

    International Nuclear Information System (INIS)

    Tsai, M.S.

    1985-01-01

    Operating regions for primary coolant feed and bleed cooling are developed for the Midland Plant using core decay heat, the high-pressure injection (HPI) system capacity, and flow rate relief through the power-operated relief valve (PORV). This mode of cooling is used for accident scenarios in which the normal core cooling means of a nuclear power plant is lost because of loss of water inventory in the steam generators. The HPI flow is based on the capacities of one and two pumps. Saturated steam, saturated water, and subcooled water are considered to be possible states of the fluid being relieved through the PORV. In estimating the PORV relief rate, flow equations are derived from the Electric Power Research Institute test data obtained from the same model and size valve that is used in the Midland Plant. For easy reference by operators, the operating region is displayed on a plane of reactor coolant system pressure and temperature. The technique developed for the Midland Plant provides a convenient method for examining the feed and bleed cooling capability for a nuclear power plant that employs a pressurized water reactor system

  20. 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)

  1. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  2. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    International Nuclear Information System (INIS)

    Solyany, V.I.; Bibilashvili, Yu.K.; Sukhanov, G.I.; Pimenov, Yu.V.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-01-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness. (author)

  3. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Solyany, V I; Bibilashvili, Yu K; Sukhanov, G I; Pimenov, Yu V [Vsesoyuznyj Nauchno-Issledovatel' skij Inst. Neorganicheskikh Materialov, Moscow (USSR); Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-12-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness.

  4. The effect of outflowing water coolant with supercritical parameters on a barrier

    Directory of Open Access Journals (Sweden)

    Alekseev Maksim

    2017-01-01

    Full Text Available The outflow of supercritical coolant with different initial parameters and its impact on the barrier have been numerically simulated. Spatial and axial distributions of pressure and steam quality are presented. The force acting on the barrier at different parameters of the outflow has been calculated.

  5. A {open_quotes}zero waste{close_quotes} coolant management strategy

    Energy Technology Data Exchange (ETDEWEB)

    Kennicott, M.A.

    1994-04-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant`s Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant`s liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study`s goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a {open_quotes}zero waste{close_quotes} machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs.

  6. Break-up and quench behavior of molten material in coolant

    International Nuclear Information System (INIS)

    Abe, Y.; Kizu, T.; Arai, T.; Nariai, H.; Chitose, K.; Koyama, K.

    2003-01-01

    In a Core Disruptive Accident (CDA) of a Fast Breeder Reactor, the Post Accident Heat Removal(PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. The material, being fragmented while solidification and forming debris bed, will be cooled in the coolant. In the experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation and boiling phenomena during PAHR in CDA. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The experimental results are compared with the existing theories. Consequently, the marginal wavelength on the surface of a water jet is close to the value estimated based on the Rayleigh-Taylor instability. Moreover, the fragmented droplet diameter obtained from the interaction of molten material and water is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass of the molten particle with some appropriate heat transfer model

  7. The calculation of coolant leak rate through the cracks using RELAP5 code

    International Nuclear Information System (INIS)

    Krungeleviciute, V.; Kaliatka, A.

    2001-01-01

    For reason to choose method of leak detection first of all it is necessary to perform evaluating thermal-hydraulic calculations. These calculations allow to determine flow rate of discharged coolant. For coolant leak rate calculations through possible cracks in Ignalina NPP pipes SQUIRT and RELAP5 thermal-hydraulic codes were used. SQUIRT is well known as computer program that predicts the leakage for cracked pipes in NPP. As this code calculates only water (at subcooled or saturated conditions) leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. For model validation comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows RELAP5 code model suitability for calculations of leak through through-wall cracks in pipes. (author)

  8. Development of Hplc Techniques for the Analysis of Trace Metal Species in the Primary Coolant of a Pressurised Water Reactor.

    Science.gov (United States)

    Barron, Keiron Robert Philip

    Available from UMI in association with The British Library. The need to monitor corrosion products in the primary circuit of a pressurised water reactor (PWR), at a concentration of 10pg ml^{-1} is discussed. A review of trace and ultra-trace metal analysis, relevant to the specific requirements imposed by primary coolant chemistry, indicated that high performance liquid chromatography (HPLC), coupled with preconcentration of sample was an ideal technique. A HPLC system was developed to determine trace metal species in simulated PWR primary coolant. In order to achieve the desired detection limit an on-line preconcentration system had to be developed. Separations were performed on Aminex A9 and Benson BC-X10 analytical columns. Detection was by post column reaction with Eriochrome Black T and Calmagite Linear calibrations of 2.5-100ng of cobalt (the main species of interest), were achieved using up to 200ml samples. The detection limit for a 200ml sample was 10pg ml^{-1}. In order to achieve the desired aim of on-line collection of species at 300^circ C, the use of inorganic ion-exchangers is essential. A novel application, utilising the attractive features of the inorganic ion-exchangers titanium dioxide, zirconium dioxide, zirconium arsenophosphate and pore controlled glass beads, was developed for the preconcentration of trace metal species at temperature and pressure. The performance of these exchangers, at ambient and 300^ circC was assessed by their inclusion in the developed analytical system and by the use of radioisotopes. The particular emphasis during the development has been upon accuracy, reproducibility of recovery, stability of reagents and system contamination, studied by the use of radioisotopes and response to post column reagents. This study in conjunction with work carried out at Winfrith, resulted in a monitoring system that could follow changes in coolant chemistry, on deposition and release of metal species in simulated PWR water loops. On

  9. 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

  10. 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.)

  11. 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

  12. Small break loss of coolant accident analysis of advanced PWR plant designs utilizing DVI line venturis

    International Nuclear Information System (INIS)

    Kemper, Robert M.; Gagnon, Andre F.; McNamee, Kevin; Cheung, Augustine C.

    1995-01-01

    The Westinghouse Advanced Passive and evolutionary Pressurizer Water Reactors (i.e. AP600 and APWR) incorporate direct vessel injection (DVI) of emergency core coolant as a means of minimizing the potential spilling of emergency core cooling water during a loss of coolant accident (LOCA). As a result, the most limiting small break LOCA (SBLOCA) event for these designs, with respect core inventory makeup capability, is a postulated double ended rupture of one of the DVI lines. This paper presents the results of a design optimization study that examines the installation of a venturi in the DVI line as a means of limiting the reactor coolant lost from the reactor vessel. The comparison results demonstrate that by incorporating a properly sized venturi in the DVI line, core uncovery concerns as a result of a DVI line break can be eliminated for both the AP600 and APWR plants. (author)

  13. 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

  14. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  15. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

    2011-01-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  16. 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)

  17. Analysis of a hot-leg small break loss-of-coolant accident in a three-loop westinghouse pressurized water reactor plant

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Clements, T.B.

    1985-01-01

    The RETRAN-02 computer code was used to perform a best-estimate analysis of a 7.52-cm-diam hotleg break in a three-loop Westinghouse pressurized water reactor. This break size produced a net primary coolant mass depletion through the early portion of the transient. The primary system started to refill only after the accumulator valves opened. As the primary system refilled, there were extreme temperature differentials around the system with cold, denser fluid collecting at the lower elevations and two-phase fluid at higher elevations

  18. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  19. 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

  20. The measurement of moisture in nuclear coolant gases. Experience and new developments within the CEGB

    International Nuclear Information System (INIS)

    Hiorns, D.S.; Stallard, M.D.

    1982-06-01

    Humidity measurements on nuclear reactor coolant gases are required for many reasons, for instance to detect and locate water leakages from the boilers and to monitor the chemical composition of the coolant for normal operation and during commissioning and maintenance, and the assessment of the performance of driers. For leak detection sensitivity and rapid response are important whereas for routine monitoring drift may be a prime criterion. Some coolants include methane for chemical control that decomposes to water so the levels of moisture that normally exist in the coolant vary with reactor types. No one measurement technique or hygrometer system currently available can be expected to meet all these requirements and in the CEGB it has proved necessary to concentrate on finding a solution to each separately. This paper reviews CEGB experience with numerous hygrometer systems that have been used on both Magnox and AGR coolant gases and highlights some of the problems that have arisen. New development work in the important and demanding area of boiler leak detection and location is also included where a number of new approaches are being investigated, including:- (i) An improved 'first-up' system based on coulometric moisture analysers. (ii) A differential system based on the piezo-electric hygrometer. (iii) A microprocessor controlled system using capacitance probes. Whilst work on the development of new instruments and the design of approved systems is proceeding at different locations within the CEGB it is being reported, together with the latest operational experiences, through the CEGB Chemical Measuring Instruments Advisory Group, on whose behalf this paper is presented. (author)

  1. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  2. Water chemistry

    International Nuclear Information System (INIS)

    Hofstetter, K.J.; Baston, V.F.

    1986-01-01

    Prior to the accident, the coolants in the primary and secondary systems were within normal chemistry specifications for an operating pressurized water reactor with once-through steam generators. During and immediately after the accident, additional boric acid and sodium hydroxide were added to the primary coolant for control of criticality and radioiodine solubility. A primary to secondary leak developed contaminating the water in one steam generator. For about 5 years after the accident, the primary coolant was maintained at 3800 +. 100 ppm boron and 1000 +. 100 ppm sodium concentrations. Dissolved oxygen was maintained 7.5, corrosion caused by increased dissolved oxygen levels (up to 8 ppm) and higher chloride ion content (up to 5 ppm) is minimized. Chemical control of dissolved oxygen was discontinued and the coolant was processed. Prior to removal of the reactor vessel head, the boron concentration in the coolant was increased to ≅ 5000 ppm to support future defueling operations. Decontamination of the accident generated water is described in terms of contaminated water management. In addition, the decontamination and chemical lay-up conditions for the secondary system are presented along with an overview of chemical management at TMI-2

  3. Measurement of actinides in samples from effluent air, primary coolant and effluent water of nuclear power stations in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Winkler, R.; Hoetzl, H.; Rosner, G.

    1977-01-01

    Since the middle of 1973 the alpha radioactivity of a number of aerosol filters from the stack monitoring systems of some nuclear power stations, of water effluent samples from all german nuclear power stations and of samples from the primary coolant water of one nuclear power reactor was measured. Essentially, the following procedures of sample preparation for alpha spectrometry of the samples in large area gridded ionization chambers were used; cold ashing of the aerosol samples in 'excited' oxygen, coprecipitation of the alpha emitters from the effluent water samples with iron hydroxide and subsequent cold ashing of the precipitate, and evaporation of the samples from the primary cycle on stainless steel plates. The following transuranium nuclides, or some of them, were found in the samples of the primary coolant and in several aerosol filter samples: Pu-239/240, Pu-238 and/or Am-241, Cm-242 and Cm-244. Cm-242 contributes most to the alpha radioactivity in fresh samples. In the effluent water samples Cm-242, Pu-239/240 and Pu-238 and/or Am-241 were identified in some cases, in one case also Cm-244. Detection limits of the procedures used for the analysis of the above stated transuranium nuclides were in the order of 0,1 fCi per m 3 for the aerosol samples and of 0.2 pCi per 1 for the liquid samples. For the effluent air and water samples in most cases specific activities near the detection limit or somewhat higher were found. On the basis of the measurements, an estimation of the annual actinides releases from nuclear power stations in the Federal Republic of Germany is given

  4. 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

  5. Upper plenum dump during reflood in PWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Sudo, Yukio; Griffith, Peter.

    1981-01-01

    Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

  6. System and method for determining coolant level and flow velocity in a nuclear reactor

    Science.gov (United States)

    Brisson, Bruce William; Morris, William Guy; Zheng, Danian; Monk, David James; Fang, Biao; Surman, Cheryl Margaret; Anderson, David Deloyd

    2013-09-10

    A boiling water reactor includes a reactor pressure vessel having a feedwater inlet for the introduction of recycled steam condensate and/or makeup coolant into the vessel, and a steam outlet for the discharge of produced steam for appropriate work. A fuel core is located within a lower area of the pressure vessel. The fuel core is surrounded by a core shroud spaced inward from the wall of the pressure vessel to provide an annular downcomer forming a coolant flow path between the vessel wall and the core shroud. A probe system that includes a combination of conductivity/resistivity probes and/or one or more time-domain reflectometer (TDR) probes is at least partially located within the downcomer. The probe system measures the coolant level and flow velocity within the downcomer.

  7. The upgrade of intense pulsed neutron source (IPNS) through the change of coolant and reflector

    CERN Document Server

    Baek, I C; Iverson, E B

    2002-01-01

    The current intense pulsed neutron source (IPNS) depleted uranium target is cooled by light water. The inner reflector material is graphite and the outer reflector material is beryllium. The presence of H sub 2 O in the target moderates neutrons and leads to a higher absorption loss in the target than is necessary. D sub 2 O coolant in the small quantities required minimizes this effect. We have studied the possible improvement in IPNS beam fluxes that would result from changing the coolant from H sub 2 O to D sub 2 O and the inner reflector from graphite to beryllium. Neutron intensities were calculated for directions normal to the viewed surface of each moderator for four different cases of combinations of target coolant and reflector materials. The simulations reported here were performed using the MCNPX (version 2.1.5) computer program. Our results show that substantial gains in neutron beam intensities can be achieved by appropriate combination of target coolant and reflector materials. The combination o...

  8. Correlations between the electrochemical behaviour and surface film composition of TZM alloy exposed to divertor water coolant environments

    International Nuclear Information System (INIS)

    Maday, M.-F.; Giorgi, R.; Dikonimos-Makris, T.

    1997-01-01

    X-ray photoelectron spectroscopy (XPS) has been carried out on TZM alloy surfaces after short and long immersion tests in high temperature (250 C) aqueous environments simulating possible fusion reactor coolant conditions during operation. Phase identification by XPS was used in connection with the open circuit potential trends to suggest plausible hypotheses about TZM corrosion behaviour in the various chemical environments considered in this study. It was proposed that exposure of TZM to oxidizing water conditions produced poorly protective layers, which consist essentially of low (IV) and intermediate (V) valency Mo oxides/hydroxides. Conversely the results obtained in deaerated and reducing water conditions suggested that barrier films could develop in these environments: the phases exhibit a bilayered structure and consisted of an inner tetravalent Mo oxide/hydroxide and an outer hexavalent Mo oxide. The protective properties of such layers were attributed to the hexavalent Mo species. (orig.)

  9. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  10. Application of fuzzy set theory for safety culture and safety management assessment of Kartini research reactor

    International Nuclear Information System (INIS)

    Syarip; Hauptmanns, U.

    2000-01-01

    The safety culture status of nuclear power plant is usually assessed through interview and/or discussions with personnel and management in plant, and an assessment of the pertinent documentation. The approach for safety culture assessment described in IAEA Safety Series, make uses of a questionnaire composed of questions which require 'Yes' or 'No' as an answer. Hence, it is basically a check-list approach which is quite common for safety assessments in industry. Such a procedure ignores the fact that the expert answering the question usually has knowledge which goes far beyond a mere binary answer. Additionally, many situations cannot readily be described in such restricted terms. Therefore, it was developed a checklist consisting of questions which are formulated such that they require more than a simple 'yes' or 'no' as an answer. This allows one to exploit the expert knowledge of the analyst appropriately by asking him to qualify the degree of compliance of each of the topics examined. The method presented has proved useful in assessing the safety culture and quality of safety management of the research reactor. The safety culture status and the quality of safety management of Kartini research reactor is rated as 'average'. The method is also flexible and allows one to add questions to existing areas or to introduce new areas covering related topics

  11. In-core failure of the instrumented BWR rod by locally induced high coolant temperature

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki

    1985-12-01

    In the BWR type light water loop instrumented in HBWR, a current BWR type fuel rod pre-irradiated up to 5.6 MWd/kgU was power ramped to 50 kW/m. During the ramp, the diameter of the rod was expanded significantly at the bottom end. The behaviour was different from which caused by pellet-cladding interaction (PCI). In the post-irradiation examination, the rod was found to be failed. In this paper, the cause of the failure was studied and obtained the followings. (1) The significant expansion of the rod diameter was attributed to marked oxidation of cladding outer diameter, appeared in the direction of 0 0 -180 0 degree with a shape of nodular. (2) The cladding in the place was softened by high coolant temperature. Coolant pressure, 7MPa intruded the cladding into inside chamfer void at pellet interface. (3) At the place of the significant oxidation, an instrumented transformer was existed and the coolant flow area was very little. The reduction of the coolant flow was enhanced by the bending of the cladding which was caused in pre-irradiation stage. They are considered to be a principal cause of local closure of coolant flow and resultant high temperature in the place. (author)

  12. An evaluation of debris mobility within a PWR reactor coolant system during the recirculation mode

    International Nuclear Information System (INIS)

    Andreychek, T.S.

    1987-01-01

    To provide for the long-term cooling of the nuclear core of a Pressurized Water Rector (PWR) following a hypothetical Loss-of-Coolant Accidnet (LOCA), water is drawn from the containment sump and pumped into the reactor coolant system (RCS). It has been postulated that debris from the containment, such as dirt, sand, and paint from containment walls and in-containment equipment, could be carried into the containment sump due to the action of the RCS coolant that escapes from the breach in the piping and then flows to the sump. Once in the sump, this debris could be pumped into the Safety Injection System (SIS) and ultimately the RCS itself, causing the performance of the SIS to be degraded. Of particular interest is the potential for core blockage that may occur due to debris transport into the core region by the recirculating flow. This paper presents a method of evaluating the potential for debris from the sump to form core blockages under recirculating flow conditions following a hypothetical LOCA for a PWR

  13. Introduction to the modified TROI test facility for fuel coolant interaction under a submerged reactor vessel

    International Nuclear Information System (INIS)

    Na, Young Su; Hong, Seong-Wan; Song, Jin Ho; Hong, Seong-Ho

    2014-01-01

    The molten Fuel-Coolant Interaction (FCI) can threaten the integrity of the reactor cavity under a severe accident. A steam explosion can be occurred by the rapid energy transfer in the high-temperature corium melt jet penetrating into water, which makes the dynamic load applying to the surrounding structure. Before a steam explosion, the corium melt jet breaks into small-sized particles, and the steam is generated continuously by the film boiling on the hot surface of the melt contacting with water. The premixing phase consisting of the corium melt, water, and steam can determine the intensity of the steam explosion. Unfortunately, the previous experimental studies on the FCI phenomena have carried out under a free fall of the corium melt jet in a gas phase before interacting with water. The previous TROI (Test for Real cOrium Interaction with water) test facility, that is a well-known test facility for the FCI phenomena in the world, has observed a steam explosion under a free fall of a corium melt jet in a gas phase before contacting a coolant since 2000, which is changing to simulate the FCI phenomena under a submerged reactor vessel. This study introduces the modified TROI test facility as shown in Fig. 1 and the considerations for the experiment with success. The previous TROI test facility, that has observed the molten Fuel-Coolant Interaction (FCI) with a free fall of the prototypic corium melt in a gas phase before contacting a coolant, was modified to simulate the FCI phenomena under a submerged reactor vessel for the assessment of the In-Vessel Retention (IVR) concept, i.e., without a free-fall distance of the corium melt before contacting water. The superheated prototypic corium melt created by the cold crucible melting method moves on a releasing valve newly installed just above the water level in the interaction vessel. The corium melt will stay on a releasing valve in less than 0.2 seconds to reduce heat loss for preventing the solidification, and

  14. Membrane systems and their use in nuclear power plants. Treatment of primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Kus, Pavel; Bartova, Sarka; Skala, Martin; Vonkova, Katerina [Research Centre Rez, Husinec-Rez (Czech Republic). Technological Circuits Innovation Dept.; Zach, Vaclav; Kopa, Roman [CEZ a.s., Temelin (Czech Republic). Nuclear Power Plant Temelin

    2016-03-15

    In nuclear power plants, drained primary coolant containing boric acid is currently treated in the system of evaporators and by ion exchangers. Replacement of the system of evaporators by membrane system (MS) will result in lower operating cost mainly due to lower operation temperature. In membrane systems the feed primary coolant is separated into two output streams: retentate and permeate. Retentate stream consists of the concentrated boric acid solution together with other components, while permeate stream consists of purified water. Results are presented achieved by testing a pilot-plant unit of reverse osmosis in nuclear power plant (NPP) Temelin.

  15. EDF PWRs primary coolant purification strategies

    International Nuclear Information System (INIS)

    Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

    2012-09-01

    In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

  16. 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

  17. 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

  18. PEMODELAN KOLIMATOR DI RADIAL BEAM PORT REAKTOR KARTINI UNTUK BORON NEUTRON CAPTURE THERAPY

    Directory of Open Access Journals (Sweden)

    Bemby Yulio Vallenry

    2015-03-01

    Full Text Available Salah satu metode terapi kanker adalah Boron Neutron Capture Therapy (BNCT. BNCT memanfaatkan tangkapan neutron oleh 10B yang terendapkan pada sel kanker. Keunggulan BNCT dibandingkan dengan terapi radiasi lainnya adalah tingkat selektivitas yang tinggi karena tingkatannya adalah sel. Pada penelitian ini dilakukan pemodelan kolimator di radial beamport reaktor Kartini sebagai dasar pemilihan material dan manufature kolimator sebagai sumber neutron untuk BNCT. Pemodelan ini dilakukan dengan simulasi menggunakan perangkat lunak Monte Carlo N-Particle versi 5 (MCNP 5. MCNP 5 adalah suatu paket program untuk memodelkan sekaligus menghitung masalah transpor partikel dengan mengikuti sejarah hidup neutron semenjak lahir, bertranspor pada bahan hingga akhirnya hilang karena mengalami reaksi penyerapan atau keluar dari sistem. Pemodelan ini menggunakan variasi material dan ukurannya agar menghasilkan nilai dari tiap parameter-parameter yang sesuai dengan rekomendasi I International Atomic Energy Agency (IAEA untuk BNCT, yaitu fluks neutron epitermal (Фepi > 9 n.cm-2.s-1, rasio antara laju dosis neutron cepat dan fluks neutron epitermal (Ḋf/Фepi 0,7. Berdasarkan hasil optimasi dari pemodelan ini, material dan ukuran penyusun kolimator yang didapatkan yaitu 0,75 cm Ni sebagai dinding kolimator, 22 cm Al sebagai moderator dan 4,5 cm Bi sebagai perisai gamma. Keluaran berkas radiasi yang dihasilkan dari pemodelan kolimator radial beamport yaitu Фepi = 5,25 x 106 n.cm-2s-1, Ḋf/Фepi =1,17 x 10-13 Gy.cm2.n-1, Ḋγ/Фepi = 1,70 x 10-12 Gy.cm2.n-1, Фth/Фepi = 1,51 dan J/Фepi = 0,731. Berdasarkan penelitian ini, hasil optimasi 5 parameter sebagai persyaratan kolimator untuk BNCT yang keluar dari radial beam port tidak sepenuhnya memenuhi kriteria yang direkomendasikan oleh IAEA sehingga perlu dilakukan penelitian lebih lanjut agar tercapainya persyaratan IAEA. Kata kunci: BNCT, radial beamport, MCNP 5, kolimator   One of the cancer therapy methods is

  19. Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Callow, R.A.

    1989-01-01

    The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

  20. 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)

  1. 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)

  2. Water chemistry control in HTTR

    International Nuclear Information System (INIS)

    Sekita, Kenji; Furusawa, Takayuki; Emori, Koichi; Kuroha, Misao; Hayakawa, Masato; Ohuchi, Hiroshi; Ishii, Taro

    2008-08-01

    A carbon steel is used for the main material for the components and pipings of the pressurized water cooling system etc. that are the reactor cooling system of the HTTR. Water quality is managed by using the hydrazine in the coolant of the water cooling system to prevent corrosion of the components and deoxidize the coolant. Also, regular analysis is carried out for the confirmation of the water quality. The following results were obtained through the water quality analysis. (1) In the pressurized water cooling system, the coolant temperature rises higher due to the heat removal of the primary coolant. So, the ammonia was formed in the thermal decomposition of the hydrazine. The electric conductivity increased, while the concentration of the hydrazine decreased, there was no problem as the plan it. (2) Thermal decomposition of the hydrazine was not occurred in the auxiliary water cooling system and vessel cooling system because of the coolant temperature was low. (3) An indistinct procedure is clarified and procedure of water quality analysis was established in the HTTR. (4) It is assumed that the corrosion of the components in these water cooling system hardly occurred from measurement results of dissolved oxide and chloride ion. Thus, the water quality was managed enough. (author)

  3. 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.)

  4. 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.)

  5. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  6. Cooling Characteristics of the V-1650-7 Engine. II - Effect of Coolant Conditions on Cylinder Temperatures and Heat Rejection at Several Engine Powers

    Science.gov (United States)

    Povolny, John H.; Bogdan, Louis J.; Chelko, Louis J.

    1947-01-01

    An investigation has been conducted on a V-1650-7 engine to determine the cylinder temperatures and the coolant and oil heat rejections over a range of coolant flows (50 to 200 gal/min) and oil inlet temperatures (160 to 2150 F) for two values of coolant outlet temperature (250 deg and 275 F) at each of four power conditions ranging from approximately 1100 to 2000 brake horsepower. Data were obtained for several values of block-outlet pressure at each of the two coolant outlet temperatures. A mixture of 30 percent by volume of ethylene glycol and 70-percent water was used as the coolant. The effect of varying coolant flow, coolant outlet temperature, and coolant outlet pressure over the ranges investigated on cylinder-head temperatures was small (0 deg to 25 F) whereas the effect of increasing the engine power condition from ll00 to 2000 brake horsepower was large (maximum head-temperature increase, 110 F).

  7. Water sampling device for detecting fuel failure

    International Nuclear Information System (INIS)

    Masubuchi, Yukio.

    1997-01-01

    A notched portion is formed at the lower end of an outer cap, and an extensible air bag is disposed being in contact with the inner side of the notched portion. A compressed air is sent into the outer gap through an air supply pipe to urge coolants thereby lowering the water level. A portion of the compressed air gets out of the outer gap from the notched portion, and if air bubbles are observed on the surface of coolants in a pressure vessel of a reactor, the outer cap is confirmed to be attached to the upper lattice plate. Compressed air is supplied to the air bag to close the notched portion. Then, coolants are sucked from a water level confirmation pipe. The level of coolants is further lowered, and the compressed air is sucked from the water level confirmation pipe instead of the coolants. Then, the level of the coolants at the inner side of the inner cap is confirmed to be made lower than the upper end of the channel box of a reactor fuel assembly. Then, coolants in the channel box are sampled, as a specimen water, through a water sampling pipe. (I.N.)

  8. Theoretical study on loss of coolant accident of a research reactor

    International Nuclear Information System (INIS)

    Lee, Kwon-Yeong; Kim, Wan-Soo

    2016-01-01

    Highlights: • A theoretical model of siphon breaking phenomena was developed. • A general formula using Chisholm coefficient B was proposed. • The safety requirements regarding a loss of coolant accident of research reactors could be found out. - Abstract: Under the design conditions of a research reactor, the siphon phenomenon induced by pipe rupture can cause continuous efflux of water. In order to prevent water efflux, an additional facility is necessary. A siphon breaker is a type of safety facility that can resist the loss of coolant effectively. However, analysis of siphon breaking is complex since it comprises two-phase flow and there are many inputs to be considered. For this reason, we analyzed the experimental results to develop a theoretical model of siphon breaking phenomena. Developed model is based on fluid mechanics and Chisholm model. From Bernoulli’s equation, the velocity and quantity as well as undershooting height, water level, pressure, friction coefficient, and factors related to the two-phase flow could be calculated. The Chisholm model, which is able to analyze the two-phase flow, can predict the results in a manner similar to those obtained from a real-scale experiment, and a general formula using Chisholm coefficient B was proposed in this study. Also, we verified the theoretical model and concluded that it is possible to analyze the siphon breaking. Moreover, the design conditions that can satisfy the safety requirements regarding a loss of coolant accident of research reactors could be found out by using the theoretical model. In conclusion, we propose the theoretical model which can analyze the siphon breaking as real, and it is helpful not only to analyze but also to design the siphon breaker.

  9. 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.

  10. 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

  11. 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.

  12. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  13. Liquid metal reactor development -Studies on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author)

  14. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  15. Thermal-hydraulic modeling of nanofluids as the coolant in VVER-1000 reactor core by the porous media approach

    International Nuclear Information System (INIS)

    Jahanfarnia, G.; Zarifi, E.; Veysi, F.

    2013-01-01

    The aim of this study was to perform a thermal-hydraulic analysis of nanofluids as coolant in the Bushehr VVER-1000 reactor core using the porous media approach. Water-based nanofluids containing various volume fractions of Al 2 O 3 and TiO 2 nanoparticles were analyzed. The conservation equations were discretized by the finite volume method and solved by numerical methods. To validate the approaches applied in this study, the results of the model and COBRA-EN code were compared for pure water. The achieved results show that the temperature of the coolant increases with the concentration of the nanoparticles. (authors)

  16. Measuring device for the coolant flowrate in a reactor core

    International Nuclear Information System (INIS)

    Sawa, Toshihiko.

    1983-01-01

    Purpose: To improve the operation performance by enabling direct and accurate measurement for the reactor core recycling flowrate. Constitution: A control rod guide is disposed to the upper end of a control rod drive mechanism housing passing through the bottom of a reactor pressure vessel and it is inserted into the through hole of a reactor core support plate. A water flow passage is formed through the reactor core support plate for the flowrate measurement of coolants recycled within the reactor core. The static pressure difference between the upper and the lower sides of the reactor core support plate is measured by a pressure difference detector of a pressure difference measuring mechanism, and an output signal from the pressure different detector is inputted to a calculation means, in which the amount of the coolants passing through the water flow passage is calculated based on the output signal corresponding to the pressure difference. Then, the total recycling flowrate in the reactor core is determined in the calculation means based on the relation between the measured flowrate and a predetermined total reactor core recycling flowrate. (Horiuchi, T.)

  17. Residual-stresses in austenitic stainless-steel primary coolant pipes and welds of pressurized-water reactors

    International Nuclear Information System (INIS)

    Faure, F.; Leggatt, R.H.

    1996-01-01

    Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between -250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching -300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa. (author)

  18. An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

    International Nuclear Information System (INIS)

    Murphy, J.G.; Corradini, M.L.

    1997-01-01

    The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

  19. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    International Nuclear Information System (INIS)

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

  20. Sound velocity in the coolant of boiling nuclear reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

  1. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  2. 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

  3. 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

  4. An improved water cooled nuclear reactor and pressuriser assembly

    International Nuclear Information System (INIS)

    Gardner, F.J.; Strong, R.

    1991-01-01

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

  5. The effect of coolants on the performance of magnetic micro-refrigerators.

    Science.gov (United States)

    Silva, D J; Bordalo, B D; Pereira, A M; Ventura, J; Oliveira, J C R E; Araújo, J P

    2014-06-01

    Magnetic refrigeration is an alternative cooling technique with envisaged technological applications on micro- and opto-electronic devices. Here, we present a magnetic micro-refrigerator cooling device with embedded micro-channels and based on the magnetocaloric effect. We studied the influence of the coolant fluid in the refrigeration process by numerically simulating the heat transfer processes using the finite element method. This allowed us to calculate the cooling power of the device. Our results show that gallium is the most efficient coolant fluid and, when used with Gd5Si2Ge2, a maximum power of 11.2 W/mm3 at a working frequency of -5 kHz can be reached. However, for operation frequencies around 50 Hz, water is the most efficient fluid with a cooling power of 0.137 W/mm3.

  6. 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.

  7. 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)

  8. 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

  9. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  10. Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

    International Nuclear Information System (INIS)

    Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

    1976-04-01

    The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

  11. 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)

  12. Application of radcal gamma thermometer assemblies for coolant monitoring in Ringhals W-PWRs

    International Nuclear Information System (INIS)

    Smith, R.D.; Romslo, K.; Moen, Oe.

    1982-07-01

    A study has been carried out investigating how Radcal Gamma Thermometers (RGTs) can be used for coolant inventory and core cooling monitoring in the Ringhals Westinghouse PWRs. The study concludes that two types of RGT rods would be required to come up with a complete solution covering both coolant inventory and core cooling monitoring. Above-core RGT rods will be installed in the guide tubes housing the outlet thermocouples. The Above-Core RGT rod is designed with 8 sensors where 4 are located in the upper head and 4 in the plenum. This rod will give an early warning about loss of coolant or void formation in the space from top of fuel to the reactor lid. A ninth thermocouple in this rod will measure the core outlet temperature as did the thermocouple the RGT rod replaced. The Above-Core RGT rods will give an early warning about approach to Inadequate Core Cooling (ICC) by measuring the collapsed water level inside the thermocouple guide tube. Four such rods are recommended per reactor. In-Core RGT rods are inserted from the seal table. These rods will give the information required for intelligent accident management in case ICC has developed. The signals obtainable from the rods will give direct information about fuel decay heat, core heat transfer conditions, core temperature and core coolant water level. The In-Core RGT rods can be used for local power monitoring during normal operation. Such a system can be shown to be economically motivated from a reactor operation point of view due to increased sensor lifetime, more accurate local power measurements, simpler physics corrections to signals, lower exposure to maintenance personnel. The signal transmission to the control room has been discussed, and ways have been indicated for presenting the information available to the operators. (Authors)

  13. Corrosion products in the coolant circuits of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1984-01-01

    The characteristics of corrosion products formed in the primary and secondary circuits of pressurized light water nuclear power plants are first briefly recalled. The problem set by the pollution of coolants and metallic surfaces is then examined. Finally, the measures of precaution to take and the possible solutions to minimize the disturbing effects of this pollution by corrosion products are presented [fr

  14. Verification of computer code for calculation of coolant radiolysis in the VVER reactor core with regard for boiling in its upper part

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, O.P.; Kabakchi, S.A. [OKB Gidropress, Podolsk, Moscow (Russian Federation)

    2010-07-01

    Code Bora for WWER coolant radiolysis calculation considering single jets boiling in the reactor core top part is developed on the basis of computer codes MOPABA-H2 (radiolysis of aqueous solutions) and SteamRad (radiolysis of vapor). Physico-chemical processes taking place in boiling core coolant are complex and diversified. Still, for the solution of certain problems their simulation can be simplified. The approach of reasonable simplification was used for development of code Bora: mathematical model assumed is purposed for simulation of phenomena only in the area of interest; the number of simulated chemical reactions and particles shall be reasonably minimum; complexity of interphase mass transfer calculation procedure shall be adequate to actually available accuracy of modeling. The analysis of new experimental initial yields of water radiolysis products data and kinetic parameters of elementary chemical reactions with their participation has been carried out. Some changes have been introduced in the mechanism of liquid water and aqueous solutions of ammonia radiolysis have been significantly revised on the basis of this analysis. Examples of the calculations provided for code Bora verification are presented. Despite of very simple simulation of interphase mass transfer, Bora allows to obtain average chemical composition of two-phase coolant at BWR core outlet with the accuracy sufficient for engineering calculations. The report also presents the results of two-phase coolant chemical composition test calculation for reactor core top part coolant boiling in pressurized water reactor. (author)

  15. The determination of magnesium in simulated PWR coolant by graphite furnace atomic absorption spectrometry

    International Nuclear Information System (INIS)

    Gatford, C.; Torrance, K.

    1988-06-01

    The determination of magnesium in simulated PWR coolant has been investigated by graphite furnace atomic absorption spectrometry with atomization from a L'vov platform. The presence of boric acid in the coolant suppresses the magnesium absorption to such an extent that removal of the boron is necessary and three variations of a methyl borate volatilization technique for the in situ removal of boron from the sample platform were investigated. This work has shown that dilution of the sample with an equal volume of acidified methanol and volatilization of the methyl borate was adequate for the determination of magnesium in coolant samples containing up to 2000 mg 1 -1 of boron. In simulated coolant samples containing 25 and 4 μg 1 -1 of magnesium, positive biases of about 2 and 0.5 μg 1 -1 were measured and these errors were considered to be due to contamination. The limit of detection in the presence of 100 and 2000 mg 1 -1 boron were 0.14 and 0.93 μg 1 -1 respectively. These performance characteristics suggest the method is completely acceptable for monitoring the chemical purity of PWR coolant and associated waters containing boric acid. If, however, more precise analyses were to be required for research purposes then any significant improvement in the above figures would require increased purity of reagents, clean-room conditions to reduce contamination and a more versatile atomic absorption spectrophotometer. (author)

  16. 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

  17. 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.

  18. 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...

  19. Numerical FEM Analyses of primary coolant system at NPP Temelin

    International Nuclear Information System (INIS)

    Junek, L.; Slovacek, M.; Ruzek, L.; Moulis, P.

    2003-01-01

    The main goal of this paper is to inform about the beginning and first steps of implementation of an aging management system at the Temelin NPP. The aging management system is important not only for achieving the current safety level but also for reaching operational reliability of a production unit equipment above the life time assumed by the original design, typically over 40 years. A method to locate the most prominent degradation regions is described. A global shell model of the primary coolant system including all loops and their components - reactor pressure vessel (RPV), steam generator (SG), main coolant pump (MCP), pressurizer, feed water and steam pipelines system is presented. The results of stress-strain analysis on the measured service parameters base are given. Validation of the results is very important and the method to compare the service measurement data with the numerical results is described. The global/local approach is mentioned and discussed. The effects of the complete global system on the individual components under monitoring are transformed into more accurate local spatial models. The local spatial models are used to analyze the gradual lifetime exhaustion of a facility during its service operation. Two spatial local models are presented, viz. feed water nozzle of SG and main coolant piping system T-brunch. The results of analysis of the local spatial models are processed by the neural network computing method, which is also described. The actual gradual damage of the material of the components under monitoring can be obtained based on the analyses performed and on the results from the neural network in combination with the knowledge of the real material characteristics. The procedures applied are included in the DIALIFE diagnostic system

  20. Water chemistry features of advanced heavy water reactor

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Experiment data report for LOFT large-break loss-of-coolant experiment L2-5

    International Nuclear Information System (INIS)

    Bayless, P.D.; Divine, J.M.

    1982-08-01

    Selected pertinent and uninterpreted data from the third nuclear large break loss-of-coolant experiment (Experiment L2-5) conducted in the Loss-of-Fluid Test (LOFT) facility are presented. The LOFT facility is a 50-MW(t) pressurized water reactor (PWR) system with instruments that measure and provide data on the system thermal-hydraulic and nuclear conditions. The operation of the LOFT system is typical of large [approx. 1000 MW(e)] commercial PWR operations. Experiment L2-5 simulated a double-ended offset shear of a cold leg in the primary coolant system. The primary coolant pumps were tripped within 1 s after the break initiation, simulating a loss of site power. Consistent with the loss of power, the starting of the high- and low-pressure injection systems was delayed. The peak fuel rod cladding temperature achieved was 1078 +- 13 K. The emergency core cooling system re-covered the core and quenched the cladding. No evidence of core damage was detected

  2. Water chemistry regimes for VVER-440 units: water chemistry influence on fuel cladding behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.

    1999-01-01

    In this lecture next problems of water chemistry influence on fuel cladding behaviour for VVER-440 units are presented: primary coolant technologies; water chemistry specification and control; fuel integrity considerations; zirconium alloys cladding corrosion (corrosion versus burn-up; water chemistry effect; crud deposition; hydrogen absorption; axial offset anomaly); alternatives for the primary coolant regimes

  3. 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

  4. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  5. Comparative assessment of thermophysical and thermohydraulic characteristics of lead, lead-bismuth and sodium coolants for fast reactors

    International Nuclear Information System (INIS)

    2002-06-01

    All prototype, demonstration and commercial liquid metal cooled fast reactors (LMFRs) have used liquid sodium as a coolant. Sodium cooled systems, operating at low pressure, are characterised by very large thermal margins relative to the coolant boiling temperature and a very low structural material corrosion rate. In spite of the negligible thermal energy stored in the liquid sodium available for release in case of leakage, there is some safety concern because of its chemical reactivity with respect to air and water. Lead, lead-bismuth or other alloys of lead, appear to eliminate these concerns because the chemical reactivity of these coolants with respect to air and water is very low. Some experts believe that conceptually, these systems could be attractive if high corrosion activity inherent in lead, long term materials compatibility and other problems will be resolved. Extensive research and development work is required to meet this goal. Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Federation in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. The need to exchange information on alternative fast reactor coolants was a major consideration in the recommendation by the Technical Working Group on Fast Reactors (TWGFRs) to collect, review and document the information on lead and lead-bismuth alloy coolants: technology, thermohydraulics, physical and chemical properties, as well as to make an assessment and comparison with respective sodium characteristics

  6. Model for cobalt 60/58 deposition on primary coolant piping in a boiling water reactor

    International Nuclear Information System (INIS)

    Dehollander, W.R.

    1979-01-01

    A first principles model for deposition of radioactive metals into the corrosion films of primary coolant piping is proposed. It is shown that the predominant mechanism is the inclusion of the radioactive species such as Cobalt 60 into the spinel structure of the corrosion film during the act of active corrosion. This deposition can occupy only a defined fraction of the available plus 2 valence sites of the spinel. For cobalt ions, this ratio is roughly 4.6 x 10 -3 of the total iron sites. Since no distinction is made between Cobalt 60, Cobalt 58, and Cobalt 59 in this process, the radioactivity associated with this inclusion is a function of the ratio of the radioactive species to the nonradioactive species in the water causing the corrosion of the pipe metal. The other controlling parameter is the corrosion rate of the pipe material. This can be a function of time, for example, and it shown that freshly descaled metal when exposed to the cobalt containing water can incorporate as much as 10 x 10 -3 cobalt ions per iron atom in the initial corrosion period. This has implications for the problem of decontaminating nuclear reactor piping. Equations and selected observations are presented without reference to any specifically identified reactor or utility, so as to protect any proprietary interest

  7. 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.)

  8. Analysis of accidental loss of pool coolant due to leakage in a PWR SFP

    International Nuclear Information System (INIS)

    Wu, Xiaoli; Li, Wei; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng

    2015-01-01

    Highlights: • Accidental loss of pool coolant due to leakage in a PWR SFP was studied using MAAP5. • The effect of emergency ventilation on the accident progression was investigated. • The effect of emergency injection on the accident progression was discussed. - Abstract: A large loss of pool coolant/water accident may be caused by extreme accidents such as the pool wall or bottom floor punctures due to a large aircraft strike. The safety of SFP under this circumstance is very important. Large amounts of radioactive materials would be easily released into the environment if a severe accident happened in the SFP, because the spent fuel pool (SFP) in a PWR nuclear power station (NPS) is often located in the fuel handing building outside the reactor containment. To gain insight into the loss of pool coolant accident progression for a pressurized water reactor (PWR) SFP, a computational model was established by using the Modular Accident Analysis Program (MAAP5). Important factors such as Zr oxidation by air, air natural circulation and thermal radiation were considered for partial and complete drainage accidents without mitigation measures. The calculation indicated that even if the residual water level was in the active fuel region, there was a chance to effectively remove the decay heat through axial heat conduction (if the pool cooling system failed) or steam cooling (if the pool cooling system was working). For sensitivity study, the effects of emergency ventilation and water injection on the accident progression were analyzed. The analysis showed that for the current configuration of high-density storage racks, it was difficult to cool the spent fuels by air natural circulation. Enlarging the space between the adjacent assemblies was a way of increasing air natural circulation flow rate and maintaining the coolability of SFP. Water injection to the bottom of the SFP helped to recover water inventory, quenching the high temperature assemblies to prevent

  9. THYDE-B1/MOD2: a computer code for analysis of small-break loss-of-coolant accidents of boiling water reactors

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Muramatsu, Ken; Kukita, Yutaka; Tasaka, Kanji

    1988-04-01

    THYDE-B1/MOD2 is a fast-running best estimate (BE) computer code to analyze thermal-hydraulic behaviors of the reactor cooling system of a boiling water reactor (BWR), mainly, during a small-break loss-of-coolant accident (SBLOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions consist of subcooled liquid, saturated mixture and saturated steam regions from the volume bottom. The regions are separated by two horizontal moving boundaries which are tracked by mass and energy balances for each region. With this three region node model, the interior of the pressure vessel can be represented by only two volumes: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous node model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SBLOCAs in which the thermal-hydraulic behavior is relatively slow and gravity controlled. The code has been improved and modified from the last version of the code, THYDE-B1/MOD1, especially in the phase separation model which is used in the mixture level calculation in the three region node model. Then, a good predictability of the code has been indicated through the comparison of calculated results with various SBLOCA test data including ROSA-III of JAERI and FIST of the General Electric Co. This report presents the code modifications and input data requirements of the THYDE-B1/MOD2 code. (author)

  10. The role of two-phase coolant in moderating fretting in nuclear steam generators

    International Nuclear Information System (INIS)

    Dyke, J.M.

    2004-01-01

    This paper expands the principal of coolant-cushioning in Nuclear Steam Generators whereby the two-phase coolant, especially the bubble film on the tube surface, moderates the vibration of coolant tubes against their supports. The current paper addresses tube bundle and anti-vibration bars (AVB) geometry issues; examines the tube bundle-coolant-AVB interfaces and examines implications for recirculation flow, AVB design and boiler size. In a T(sat) fluid, the tube surface is uniformly coating with growing bubbles whose momentum is perpendicular to the surface at first, then they are swept away by the bulk flow. The combination of this momentum, the phase change and the water film remaining on the surface, counteract the vibration energy of the tube-AVB system, reducing the likelihood of metal-to-metal contact and consequent fretting. To maximize the benefit of the cushioning effect, the following design inputs are needed: 1) the AVB-tube interface should have sufficient clearance for the T(sat) solution to operate, 2) The AVB should be wide enough to generate the necessary cushioning force, and 3) the AVB should be thin enough to be flexible and absorb some of the transferred vibration energy. Furthermore, fretting and crude deposition at the AVB-tube interface can be reduced or eliminated by reducing the number of AVBs, increasing clearances and making the AVBs limber

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1967-01-01

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

  12. 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

  13. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

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

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  14. Station blackout with reactor coolant pump seal leakage

    International Nuclear Information System (INIS)

    Evinay, A.

    1993-01-01

    The U.S. Nuclear Regulatory Commission (NRC) amended its regulations in 10CFR50 with the addition of a new section, 50.63, open-quotes Loss of All Alternating Current Power.close quotes The objective of these requirements is to ensure that all nuclear plants have the capability to withstand a station blackout (SBO) and maintain adequate reactor core cooling and containment integrity for a specified period of time. The NRC also issued Regulatory Guide (RG) 1.155, open-quotes Station Blackout,close quotes to provide guidance for meeting the requirements of 10CFR50.63. Concurrent with RG-1.155, the Nuclear Utility Management and Resources Council (NUMARC) has developed NUMARC 87-00 to address SBO-coping duration and capabilities at light water reactors. Licensees are required to submit a topical report based on NUMARC 87-00 guidelines, to demonstrate compliance with the SBO rule. One of the key compliance criteria is the ability of the plant to maintain adequate reactor coolant system (RCS) inventory to ensure core cooling for the required coping duration, assuming a leak rate of 25 gal/min per reactor coolant pump (RCP) seal in addition to technical specification (TS) leak rate

  15. 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

  16. 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

  17. Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

    Directory of Open Access Journals (Sweden)

    Istvan Farkas

    2016-08-01

    Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

  18. 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)

  19. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

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

  20. 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.

  1. 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.

  2. 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

  3. Fuel-coolant interactions in a jet contact mode

    International Nuclear Information System (INIS)

    Konishi, K.; Isozaki, M.; Imahori, S.; Kondo, S.; Furutani, A.; Brear, D.J.

    1994-01-01

    Molten fuel-coolant interactions in a jet contact mode was studied with respect to the safety of liquid-metal-cooled fast reactors (LMFRs). From a series of molten Wood's metal (melting point: 79 deg. C, density: -8400 kg/m 3 ) jet-water interaction experiments, several distinct modes of interaction behaviors were observed for various combinations of initial temperature conditions of the two fluids. A semi-empirical model for a minimum film boiling temperature criterion was developed and used to reasonably explain the different interaction modes. It was concluded that energetic jet-water interactions are only possible under relatively narrow initial thermal conditions. Preliminary extrapolation of the present results in an oxide fuel-sodium system suggests that mild interactions with short breakup length and coolable debris formation should be most likely in LMFRs. (author)

  4. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    Salaices, M.; Salaices, E.; Ovando, R.; Esquivias, J.

    2011-11-01

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  5. 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.

  6. Fluid-Structure Interaction for Coolant Flow in Research-type Nuclear Reactors

    International Nuclear Information System (INIS)

    Curtis, Franklin G.; Ekici, Kivanc; Freels, James D.

    2011-01-01

    The High Flux Isotope Reactor (HFIR), located at the Oak Ridge National Laboratory (ORNL), is scheduled to undergo a conversion of the fuel used and this proposed change requires an extensive analysis of the flow through the reactor core. The core consists of 540 very thin and long fuel plates through which the coolant (water) flows at a very high rate. Therefore, the design and the flow conditions make the plates prone to dynamic and static deflections, which may result in flow blockage and structural failure which in turn may cause core damage. To investigate the coolant flow between fuel plates and associated structural deflections, the Fluid-Structure Interaction (FSI) module in COMSOL will be used. Flow induced flutter and static deflections will be examined. To verify the FSI module, a test case of a cylinder in crossflow, with vortex induced vibrations was performed and validated.

  7. The origin and magnitude of pressures in fuel-coolant interactions

    International Nuclear Information System (INIS)

    Heer, W.; Jakeman, D.; Smith, B.L.

    1987-01-01

    A number of small scale experiments to simulate fuel coolant interaction (FCI) effects have been carried out using Freon and water. Contrary to the predictions of most current FCI models, only modest pressure transients are observed within the interaction region itself but large pressure spikes, near to or above critical Freon pressure, are seen at the boundaries of the region. Similar pressure amplification effects have been noticed in parallel experiments involving two phase mixtures. It is suggested that in both cases a water hammer type effect is the cause of the pressure spikes. These observations could form the basis of new thinking in FCI modelling. (author)

  8. Performance Investigation of Automobile Radiator Operated with ZnFe2O4 Nano Fluid based Coolant

    Directory of Open Access Journals (Sweden)

    Tripathi Ajay

    2015-01-01

    Full Text Available The cooling system of an Automobile plays an important role in its performance, consists of two main parts, known as radiator and fan. Improving thermal efficiency of engine leads to increase the engine's performance, decline the fuel consumption and decrease the pollution emissions. Water and ethylene glycol as conventional coolants have been widely used in radiators of an automotive industry 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 preparation of Zinc based nanofluids (ZnFe2O4 using chemical co-precipitation method and its application in an automotive cooling system along with mixture of ethylene glycol and water (50:50. Relevant input data, nanofluids properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nano fluid-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 base-fluid compared to ethylene glycol (i.e. base-fluid alone. It is observed that, about 78% of heat transfer enhancement could be achieved with the addition of 1% ZnFe2O4 particles in a base fluid at the Reynolds number of 84.4x103 and 39.5x103 for air and coolant respectively

  9. Study on the quench behavior of molten fuel material jet into coolant

    International Nuclear Information System (INIS)

    Abe, Yutaka; Kizu, Tetsuya; Arai, Takahiro; Nariai, Hideki; Chitose, Keiko; Koyama, Kazuya

    2004-01-01

    In a core disruptive accident (CDA) of a Fast Breeder Reactor, the post accident heat removal (PAHR) is crucial for the accident mitigation. The molten core material should be solidified in the sodium coolant in the reactor vessel. In the present experiment, molten material jet is injected into water to experimentally obtain fragments and the visualized information of the fragmentation. The distributed particle behavior of the molten material jet is observed with high-speed video camera. The distributions of the fragmented droplet diameter from the molten material jet are evaluated by correcting the solidified particles. The experimental results of the mean fragmented droplet diameter are compared with the existing theories. Consequently, the fragmented droplet diameter is close to the value estimated based on the Kelvin-Helmholtz instability. Once the particle diameter of the fragmented molten material could be known from a hydrodynamic model, it becomes possible to estimate the mass ratio of the molten particle to the total injected mass by combining an appropriate heat transfer model. The heat transfer model used in the present study is composed of the fragmentation model based on the Kelvin-Helmholtz instability. The mass ratio of the molten fragment to total mass of the melted mixed oxide fuel in sodium coolant estimated in the present study is very small. The result means that most of the molten mixed oxide fuel material injected into the sodium coolant can be cooled down under the solidified temperature, that is so called quenched, if the amount of the coolant is sufficient. (author)

  10. A Comparison of Coolant Options for Brayton Power Conversion Heat Rejection Systems

    International Nuclear Information System (INIS)

    Siamidis, John; Mason, Lee

    2006-01-01

    This paper describes potential heat rejection design concepts for Brayton power conversion systems. Brayton conversion systems are currently under study by NASA for Nuclear Electric Propulsion (NEP) and surface power applications. The Brayton Heat Rejection Subsystem (HRS) must dissipate waste heat generated by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Sodium potassium (NaK) and H2O are two coolant working fluids that have been investigated in the design of a pumped loop and heat pipe space HRS. In general NaK systems are high temperature (300 to 1000 K) low pressure systems, and H2O systems are low temperature (300 to 600 K) high pressure systems. NaK is an alkali metal with health and safety hazards that require special handling procedures. On the other hand, H2O is a common fluid, with no health hazards and no special handling procedures. This paper compares NaK and H2O for the HRS pumped loop coolant working fluid. A detailed excel analytical model, HRS O pt, was developed to evaluate the various HRS design parameters. It is capable of analyzing NaK or H2O coolant, parallel or series flow configurations, and numerous combinations of other key parameters (heat pipe spacing, diameter and radial flux, radiator facesheet thickness, fluid duct system pressure drop, system rejected power, etc.) of the HRS. This paper compares NaK against water for the HRS coolant working fluid with respect to the relative mass, performance, design and implementation issues between the two fluids

  11. Analysis of the core reflooding of a PWR reactor under a loss-of-coolant postulated accident

    International Nuclear Information System (INIS)

    Austregesilo Filho, H.

    1978-12-01

    The main purpose of this work is to analyse the termohydraulic behaviour of emergency cooling water, during reflooding of a PWR core submitted to a postulated loss-of-coolant accident, with the scope of giving the boundary conditions needed to verify fuel element and containment integrity. The analytical model presented was applied to the simulation of Angra I core reflooding phase, after a double-ended break between pressure vessel and discharge of one of the main coolant pumps. For this accident, with a discharge coefficient of C sub(D) = 0.4, the highest peak cladding temperature is expected. (author) [pt

  12. 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

  13. Knock-Limited Power Outputs from a CFR Engine Using Internal Coolants. 3; Four Alkyl Amines, Three Alkanolamines, Six Amides, and Eight Heterocyclic Compounds

    Science.gov (United States)

    Imming, Harry S.; Bellman, Donald R.

    1947-01-01

    An investigation of the antiknock effectiveness of various additive-water solutions when used as internal coolants has been conducted at the NACA Cleveland laboratory. Nine compounds have been previously run in a CFR engine and the results are presented. In an effort to find a good anti-knock-coolant additive with more desirable physical properties than those of the nine compounds previously investigated, water solutions of four alkyl amines, three alkanolamines, six amides, and eight heterocyclic compounds were investigated and the results are presented.

  14. 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)

  15. 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)

  16. Determination of boron as boric acid by automatic potentiometric titration using Gran plots [in pressurized water reactor coolant

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1989-11-01

    Boron in PWR primary coolant and related waters may be determined as boric acid by titration with sodium hydroxide, using a glass electrode as a pH indicator. Earlier work has shown that this analysis can conveniently be carried out automatically with adequate precision and accuracy for routine use, although bias became apparent at the lowest concentrations tested. The latest titrators enable the titration data to be transformed mathematically to give two linear segments, before and after the end-point (Gran plots). The results are as precise as those from other titration methods (in which the end-point is found from the point of inflexion of a plot of pH against volume of titrant), but the bias at low concentrations is much reduced. This is achieved without extra time or involvement of the operator. (author)

  17. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    Science.gov (United States)

    Xie, Huaqing; Li, Yang; Yu, Wei

    2010-05-01

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2O 3, ZnO, TiO 2, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2O 3, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  18. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    International Nuclear Information System (INIS)

    Xie Huaqing; Li Yang; Yu Wei

    2010-01-01

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al 2 O 3 , ZnO, TiO 2 , and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al 2 O 3 , and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  19. Intriguingly high convective heat transfer enhancement of nanofluid coolants in laminar flows

    Energy Technology Data Exchange (ETDEWEB)

    Xie Huaqing, E-mail: hqxie@eed.sspu.c [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China); Li Yang; Yu Wei [School of Urban Development and Environmental Engineering, Shanghai Second Polytechnic University, Shanghai 201209 (China)

    2010-05-31

    We reported on investigation of the convective heat transfer enhancement of nanofluids as coolants in laminar flows inside a circular copper tube with constant wall temperature. Nanofluids containing Al{sub 2}O{sub 3}, ZnO, TiO{sub 2}, and MgO nanoparticles were prepared with a mixture of 55 vol.% distilled water and 45 vol.% ethylene glycol as base fluid. It was found that the heat transfer behaviors of the nanofluids were highly depended on the volume fraction, average size, species of the suspended nanoparticles and the flow conditions. MgO, Al{sub 2}O{sub 3}, and ZnO nanofluids exhibited superior enhancements of heat transfer coefficient, with the highest enhancement up to 252% at a Reynolds number of 1000 for MgO nanofluid. Our results demonstrated that these oxide nanofluids might be promising alternatives for conventional coolants.

  20. 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

  1. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

  2. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

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

    International Nuclear Information System (INIS)

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

    1999-11-01

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

  4. Integrated main coolant pumps for pressurized-water reactors

    International Nuclear Information System (INIS)

    Wieser, R.

    1975-01-01

    The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

  5. 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

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

    Science.gov (United States)

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

    2010-08-01

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

  7. Analysis of small break loss of coolant accident for Chinese CPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

    2016-05-15

    This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

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

    Directory of Open Access Journals (Sweden)

    A. Rama Rao

    2008-04-01

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

  9. 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)

  10. 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

  11. Probabilistic analysis of fuel pin behaviour during an eventual loss of coolant in PWR reactors

    International Nuclear Information System (INIS)

    1981-02-01

    Brief description of the development of the coolant loss incident in a pressurized water reactor and analysis of its significance for the behaviour of the fuel rods. Description of a probalistic method for estimating the effects of the accident on the fuel rods and results obtained [fr

  12. 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

  13. Device for extracting steam or gas from the primary coolant line leading from a reactor pressure vessel to a straight through boiler or from the top primary boiler chamber of a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Schatz, K.

    1982-01-01

    In such a nuclear reactor, a steam or gas cushion can form when the primary system is refilled, which can cause blocking of the natural circulation or filling of the system in the area of the hot primary coolant pipe or in the top primary boiler chamber. In order to remove such a steam or gas cushion, a ventilation pipe starting from the bend of the primary coolant line is connected to the feed pipe for introducing water into the primary system. The feed pipe is designed on the principle of the vacuum pump in the area of the opening of the ventilation pipe. There is a sub-pressure in the ventilation pipe, which makes it possible to extract the steam or gas. After mixing in the area of the opening, the steam condenses or is distributed with the gas in the primary coolant. (orig.) [de

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  15. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  16. 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

  17. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  18. Numerical evaluation of various gas and coolant channel designs for high performance liquid-cooled proton exchange membrane fuel cell stacks

    International Nuclear Information System (INIS)

    Sasmito, Agus P.; Kurnia, Jundika C.; Mujumdar, Arun S.

    2012-01-01

    A careful design of gas and coolant channel is essential to ensure high performance and durability of proton exchange membrane (PEM) fuel cell stack. The channel design should allow for good thermal, water and gas management whilst keeping low pressure drop. This study evaluates numerically the performance of various gas and coolant channel designs simultaneously, e.g. parallel, serpentine, oblique-fins, coiled, parallel-serpentine and a novel hybrid parallel-serpentine-oblique-fins designs. The stack performance and local distributions of key parameters are investigated with regards to the thermal, water and gas management. The results indicate that the novel hybrid channel design yields the best performance as it constitutes to a lower pumping power and good thermal, water and gas management as compared to conventional channels. Advantages and limitation of the designs are discussed in the light of present numerical results. Finally, potential application and further improvement of the design are highlighted. -- Highlights: ► We evaluate various gas and coolant channel designs in liquid-cooled PEM fuel cell stack. ► The model considers coupled electrochemistry, channel design and cooling effect simultaneously. ► We propose a novel hybrid channel design. ► The novel hybrid channel design yields the best thermal, water and gas management which is beneficial for long term durability. ► The novel hybrid channel design exhibits the best performance.

  19. Application of radcal gamma thermometer assemblies for core coolant monitoring in ASEA ATOM reactors with particular reference to the Barsebaeck plants

    International Nuclear Information System (INIS)

    Romslo, K.

    1982-02-01

    In this study reference designs for instrument assemblies containing RGT rods to monitor the core coolant conditions in the Barsebaeck reactors have been worked out. Four such strings would be required to satisfy the Reg. Guide 1.97 reqiurements. The signal transmission to the control room and the presentation of information to the operators have been addressed. Downcomer water level measurement is considered important in order to get an early warning about leakages. Possible ways of diversifying the existing measurement method using RGTs are mentioned, and the design of a downcomer RGT rod has been suggested. To fully comply with Reg. Guide 1.97, water level measurements above core would be required. In a conceptual way it has been shown how an RGT rod could be extended up into this region, if so required. The possibility of making an ideal core coolant monitoring system by replacing one of the structural rods (water rods) in the fuel bundle by an RGT rod is pointed out. There are foreseen, however, several practical obstacles in pursuing the idea. The present state of RGT development and further work required to get the intrument licensed as a coolant monitoring device, has been defined. (Author)

  20. Boiling water reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

    1975-01-01

    Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

  1. 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

  2. Study on B-10 consumption of PWR primary coolant during normal operation

    International Nuclear Information System (INIS)

    Liang, C.H.

    1994-01-01

    B-10 consumption under PWR primary coolant conditions has been analyzed. The result indicates its time-dependent change reacting with neutron in the normal operation. In this work, neutron energy assumed to be 4 eV; thermal neutron flux is in the range of 3 x 10 13 to 3 x 10 14 n/sec - cm 2 and the time of cycling of the primary coolant through the RCS is 8 sec. and its retention time in the core region is about 1 sec. Under this condition investigated, B-10 consumption is less than 5% at 3 x 10 13 n/sec - cm 2 thermal neutron flux, and closes to 27% at 3 x 10 14 n/sec - cm 2 by calculation at the 16th month of continuous operation. The effect of B-10 consumption on PWR primary water chemistry is also investigated. (author). 1 fig., 2 tabs., 4 refs

  3. Fuel-coolant interaction in a shock tube with initially-established film boiling

    International Nuclear Information System (INIS)

    Sharon, A.; Bankoff, S.G.

    1979-01-01

    A new mode of thermal interaction has been employed, in which liquid metal is melted in a crucible within a shock tube; the coolant level is raised to overflow the crucible and establish subcooled film boiling with known bulk metal temperature; and a pressure shock is then initiated. With water and lead-tin alloy an initial splash of metal may be obtained after the vapor film has collapsed, due primarily to thermal interaction, followed by a successive cycle of bubble growth and collapse. To obtain large interactions, the interfacial contact temperature must exceed the spontaneous nucleation temperature of the coolant. Other cutoff behavior is observed with respect to the initial system pressure and temperatures and with the shock pressure and rise time. Experiments with butanol and lead-tin alloy show only relatively mild interactions. Qualitative explanations are proposed for the different behaviors of the two liquids

  4. Design of collimator in the radial piercing beam port of Kartini reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    M Ilma Muslih A; Andang Widiharto; Yohannes Sardjono

    2014-01-01

    Studies were carried out to design a collimator which results in epithermal neutron beam for in vivo experiment of Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni (95%) as collimator wall, 15 cm thick of Al as moderator, 1 cm thick of Pb as γ-ray shielding, 1.5 cm thick of Boral as additional material, with 2 cm aperture diameter, epithermal neutron beam with maximum flux of 5.03 x 10 8 n.cm -2 .s -1 could be produced. The beam has minimum fast neutron and γ-ray components of, respectively, 2.17 x 10 -13 Gy.cm 2 .n -1 and 1.16 x 10 -13 Gy.cm 2 .n -l , minimum thermal neutron per epithermal neutron ratio of 0.12, and maximum directionality of 0.835 . It did not fully pass the IAEA's criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 10 9 n.cm -2 .s -l . Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 10 8 n.cm -2 .s -1 and fast neutron flux close to 2 x 10 -13 Gy.cm 2 .n -1 it is still feasible for BNCT in vivo experiment. (author)

  5. 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.

  6. DETERMINATION OF THE 129I IN PRIMARY COOLANT OF PWR

    Directory of Open Access Journals (Sweden)

    KE CHON CHOI

    2013-02-01

    In this report, the effect of the boron content in a pressurized-water reactor primary coolant on the separation process of 129I was examined, as was the effect of 3H on the measurement of the activity of iodine. As a result, no influence of the boron content and of the simultaneous 3H presence was found with activity concentrations of 3H lower than 50 Bq/mL, and with a boron concentration of less than 2,000 μg/mL.

  7. 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

  8. 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

  9. Pressurizing Behavior on Ingress of Coolant into Pebble Bed of Blanket of Fusion DEMO Reactor

    International Nuclear Information System (INIS)

    Daigo Tsuru; Mikio Enoeda; Masato Akiba

    2006-01-01

    Solid breeder blankets are being developed as candidate blankets for the Fusion DEMO reactor in Japan. JAEA is performing the development of the water cooled and helium cooled solid breeder blankets. The blanket utilizes ceramic breeder pebbles and multiplier pebbles beds cooled by high pressure water or high pressure helium in the cooling tubes placed in the blanket box structure. In the development of the blanket, it is very important to incorporate the safety technology as well as the performance improvement on tritium production and energy conversion. In the safety design and technology, coolant ingress in the blanket box structure is one of the most important events as the initiators. Especially the thermal hydraulics in the pebble bed in the case of the high pressure coolant ingress is very important to evaluate the pressure propagation and coolant flow behavior. This paper presents the preliminary results of the pressure loss characteristics by the coolant ingress in the pebble bed. Experiments have been performed by using alumina pebble bed (4 litter maximum volume of the pebble bed) and nitrogen gas to simulate the helium coolant ingress into breeder and multiplier pebble beds. Reservoir tank of 10 liter is filled with 1.0 MPa nitrogen. The nitrogen gas is released at the bottom part of the alumina pebble bed whose upper part is open to the atmosphere. The pressure change in the pebble bed is measured to identify the pressure loss. The measured values are compared with the predicted values by Ergun's equation, which is the correlation equation on pressure loss of the flow through porous medium. By the results of the experiments with no constraint on the alumina pebble bed, it was clarified that the measured value agreed in the lower flow rate. However, in the higher flow rate where the pressure loss is high, the measured value is about half of the predicted value. The differences between the measured values and the predicted values will be discussed from

  10. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D P; Lukas, M; Lynch, B K [Spectro Incorporated, Littleton, MA (United States)

    1998-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  11. Diesel engine coolant analysis, new application for established instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D.P.; Lukas, M.; Lynch, B.K. [Spectro Incorporated, Littleton, MA (United States)

    1997-12-31

    Rotating disk electrode (RDE) arc emission spectrometers are user` many commercial, industrial and military laboratories throughout the world to analyze millions of oil and fuel samples each year. In fact, RDE spectrometers have been used exclusively for oil and fuel analysis for so long that it has nearly been forgotten by most practitioners that when RDE spectrometers were first introduced more than 40 years ago, they were routinely used for aqueous samples as well. This presentation reviews early methods of aqueous sample analysis using RDE technology. This presentation also describes recent work to calibrate an RDE spectrometer for both water samples and for engine coolant samples which are a mixture of approximately 50 % water and 50 % ethylene or propylene glycol. Limits of detection determined for aqueous standards are comparable to limits of detection for oil standards. Repeatability of aqueous samples is comparable to the repeatability achieved for oil samples. A comparison of results for coolant samples measured by both inductively coupled plasma (ICP) and rotating disk electrode (RDE) spectrometers is presented. Not surprisingly, RDE results are significantly higher for samples containing particles larger than a few micrometers. Although limits of detection for aqueous samples are not as low as can be achieved using the more modern ICP spectrometric method or the more cumbersome atomic absorption (AA) method, this presentation suggests that RDE spectrometers may be appropriate for certain types of aqueous samples in situations where the more sensitive ICP or AA spectrometers and the laboratory environment and skilled personnel needed for them to operate are not conveniently available. (orig.) 4 refs.

  12. Description of steam condensation phenomena during the loss-of-coolant accident

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Furst, H.; Schwan, H.; Vollbrandt, J.

    1981-01-01

    Study of results from the full scale multivent pressure suppression experiment conducted by the GKSS Laboratory has developed an improved understanding of the dynamic, oscillatory steam condensation events and related loading functions which occur during the hypothetical loss-of-coolant accident in a boiling water nuclear reactor. Due to the unique measurements systems which combines both cinematic and digital data, qualified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena has been obtained

  13. Design Features of the SMART Water Chemistry

    International Nuclear Information System (INIS)

    Byung Seon Choi; Seong Hoon Kim; Juhyeon Yoon; Doo Jeong Lee; Yoon Yeong Bae; Sung Kyun Zee

    2004-01-01

    The design features for the primary water chemistry for the SMART are introduced from the viewpoint of the system characteristics and the chemical design concept. The most essential differences in water chemistry between the commercially operating PWRs and SMART are characterized by the presence of boron in the water and the operating mode of the purification system. SMART is a soluble boron free reactor, and the ammonia is used as a pH reagent. The material for SMART steam generator is also different from the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. In SMART hydrogen gas which suppresses a generation of oxidizing species by the radiolysis processes in the reactors is not added to the primary coolant, but is normally generated from the radiolysis of the ammonia as the coolant passes through the core. Ammonia is added once per shift because SMART reactor has no letdown and charging system during power operation. Because of these competing processes, the concentrations of hydrogen, nitrogen and ammonia in the primary coolant are in equilibrium, which depend on the decomposition and/or combination rate of the ammonia. The level of permissible oxygen concentration in the primary coolant can be ensured by both suppression of the water radiolysis through maintaining a high enough hydrogen concentration in the primary coolant and by a restriction of the oxygen ingress into the primary coolant with the makeup water. The ammonia chemistry in SMART reactor eliminates the need for hydrogen injection for the control of the dissolved oxygen in the primary coolant because of spontaneous generation of hydrogen and nitrogen produced by the reaction of the ammonia decomposition. (authors)

  14. 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.

  15. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  16. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  17. Calculational advance in the modeling of fuel-coolant interactions

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1982-01-01

    A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials

  18. 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)

  19. 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

  20. 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.

  1. Consequences in a long time of the forced loss of coolant in a pool type reactor

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1986-01-01

    The fuel and pool water temperatures are calculated as a function of time using unidimensional models of heat conduction and momentum conservation, to simulate the natural convection flow of the coolant. The reactor building pressure due to the pool water evaporation is calculated using a homogeneous model with thermal equilibrium. The heat loss from the three main components of the building volume (liquid water, air, and steam) to solid surfaces such as the building walls are taking into account. (Author) [pt

  2. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  3. Research on loss of coolant accident of pressurized-water reactor based on PSO algorithm

    International Nuclear Information System (INIS)

    Ma Jie; Guo Lifeng; Peng Qiao

    2012-01-01

    In order to improve the diagnosis performance of Loss of Coolant Accident (LOCA), based on Back Propagation (BP) algorithm study, a fault diagnosis network is established based on Particle Swarm Optimization (PSO) algorithm in this paper. The PSO algorithm is used to train the weights and the thresholds of neural network, which can conquer part convergence problem of BP algorithm. The test results show that the diagnosis network has higher accuracy of LOCA. (authors)

  4. 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.

  5. 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.

  6. 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

  7. 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

  8. Survey of Water Chemistry and Corrosion of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Sok; Hong, Bong Geon

    2008-06-15

    Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented.

  9. Survey of Water Chemistry and Corrosion of NPP

    International Nuclear Information System (INIS)

    Jung, Ki Sok; Hong, Bong Geon

    2008-06-01

    Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented

  10. A study on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tai; Choi, Y D; Choi, J H; Kim, T J; Jeong, K C; Kwon, S W; Kim, B H; Jeong, J Y; Park, J H; Kim, K R; Jo, B R

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H{sub 2} gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs.

  11. A study on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tai; Choi, Y. D.; Choi, J. H.; Kim, T. J.; Jeong, K. C.; Kwon, S. W.; Kim, B. H.; Jeong, J. Y.; Park, J. H.; Kim, K. R.; Jo, B. R.

    1997-08-01

    A study on safety measures of LMR coolant showed the results as follows: 1. Sodium fire characteristics. A. Sodium pool temp., gas temp., oxygen concentration calculated by flame combustion model were generally higher than those calculated by surface combustion model. B. Basic and detail designs for medium sodium fire test facility were carried out and medium sodium fire test facility was constructed. 2. Sodium/Cover gas purification technology. A. Construction and operation of calibration loop. B. Purification analysis and conceptual design of the packing for a cold trap. 3. Analysis of sodium-water reaction characteristics. We have investigated the characteristics analysis for micro and small leaks phenomena, development of the computer code for analysis of initial and quasi steady-state spike pressures to analyze large leak accident. Also, water mock-up test facility for the analysis of large leak accident phenomena was designed and manufactured. 4. Development of water leak detection technology. Detection signals were appeared when the hydrogen detector is operated to Ar-H 2 gas system. The technology for the passive acoustic detection with respect to large leakage of water into sodium media was reviewed. And water mock-up test equipment and instrument system were designed and constructed. (author). 19 refs., 45 tabs., 52 figs

  12. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  13. 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)

  14. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    International Nuclear Information System (INIS)

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  16. 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)

  17. Sudden contact of a hot liquid with a volatile coolant: instability of the created vapour film

    International Nuclear Information System (INIS)

    Pion, Agnes

    1983-01-01

    As the sudden contact of a hot body with a coolant which may evaporate, results, after some delay, in an explosive evaporation, this research thesis proposes an interpretation based on the study of the destabilization of the vapour film which forms at the surface of the hot body. The author reports the modelling of the evolution of the average thickness of the film before the explosion. The possible chemical reactions at the surface of the hot body are taken into account. A base flow is obtained which allows the calculation of the evolution of Rayleigh-Taylor instabilities which may occur at the gas-coolant interface. This study is applied to the interaction between liquid sodium and water [fr

  18. 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

  19. 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)

  20. 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

  1. THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

    International Nuclear Information System (INIS)

    Muramatsu, Ken; Akimoto, Masayuki

    1982-08-01

    THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

  2. Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A.; Subbotin, S. A.; Chibinyaev, A. V.

    2011-01-01

    Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

  3. Additional requirements for leak-before-break application to primary coolant piping in Belgium

    Energy Technology Data Exchange (ETDEWEB)

    Roussel, G. [AIB Vincotte Nuclear, Brussels (Belgium)

    1997-04-01

    Leak-Before-Break (LBB) technology has not been applied in the first design of the seven Pressurized Water Reactors the Belgian utility is currently operating. The design basis of these plants required to consider the dynamic effects associated with the ruptures to be postulated in the high energy piping. The application of the LBB technology to the existing plants has been recently approved by the Belgian Safety Authorities but with a limitation to the primary coolant loop. LBB analysis has been initiated for the Doel 3 and Tihange 2 plants to allow the withdrawal of some of the reactor coolant pump snubbers at both plants and not reinstall some of the restraints after steam generator replacement at Doel 3. LBB analysis was also found beneficial to demonstrate the acceptability of the primary components and piping to the new conditions resulting from power uprating and stretch-out operation. LBB analysis has been subsequently performed on the primary coolant loop of the Tihange I plant and is currently being performed for the Doel 4 plant. Application of the LBB to the primary coolant loop is based in Belgium on the U.S. Nuclear Regulatory Commission requirements. However the Belgian Safety Authorities required some additional analyses and put some restrictions on the benefits of the LBB analysis to maintain the global safety of the plant at a sufficient level. This paper develops the main steps of the safety evaluation performed by the Belgian Safety Authorities for accepting the application of the LBB technology to existing plants and summarizes the requirements asked for in addition to the U.S. Nuclear Regulatory Commission rules.

  4. 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...

  5. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  6. 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.

  7. Coolant channel module CCM

    International Nuclear Information System (INIS)

    Hoeld, Alois

    2007-01-01

    A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

  8. The Design of Sample Driver System for Gamma Irradiator Facility at Thermal Column of Kartini Reactor

    International Nuclear Information System (INIS)

    Suyamto; Tasih Mulyono; Setyo Atmojo

    2007-01-01

    The design and construction of sample driver system for gamma irradiator facility at thermal column of Kartini reactor post operation has been carried out. The design and construction is based on the space of thermal column and the sample speed rotation which has to as low as possible in order the irradiation process can be more homogeneity. The electrical and mechanical calculation was done after fixation the electrical motor and transmission system which will be applied. By the assumption that the maximum sample weight is 50 kg, the electric motor specification is decided due to its rating i.e. single phase induction motor, run capacitor type, 0.5 HP; 220 V; 3.61 A, CCW and CW, rotation speed 1430 rpm. To achieve the low load rotation speed, motor speed was reduced twice using the conical reduction gear with the reduction ratio 3.9 and thread reduction gear with the reduction ratio 60. From the calculation it is found that power of motor is 118.06 watt, speed rotation of load sample is 6.11 rpm due to the no load rotation of motor 1430 rpm. From the test by varying weight of load up to 75 kg it is known that the device can be operated in a good condition, both in the two direction with the average speed of motor 1486 rpm and load 6.3 rpm respectively. So that the slip is 0.268 % and 0.314 % for no load and full load condition. The difference input current to the motor during no load and full load condition is relative small i.e. 0.14 A. The safety factor of motor is 316 % which is correspond to the weight of load 158 kg. (author)

  9. 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.

  10. 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)

  11. Conservatism of loss-of-coolant accident licensing analysis compared to experimental results and best-estimate calculation

    International Nuclear Information System (INIS)

    Winkler, F.; Friedmann, P.

    1986-01-01

    The paper compares results of loss-of-coolant accident licensing analysis with experimental results and results of best-estimate calculations. The large safety margins resulting from the more realistic best-estimate results are used to show the high conservatism inherent in the licensing process of pressurized water reactors. (orig.) [de

  12. Konsep Cantik pada Iklan Cetak Majalah Kartini Tahun 2014 dan Koran Kompas Tahun 1979: Suatu Analisis Semiotik

    Directory of Open Access Journals (Sweden)

    Dian Natashia

    2015-11-01

    Full Text Available Advertisements, which can be found in daily life, do not just function as the source of the product’s information, but can also create a new connotation about social reality. This research is intended to reveal the form of the connotation about present beauty of Indonesian women that are trying to be conveyed by the advertisers through present advertisements in Kartini magazines.  This research is also revealing elements that are used by the advertisers to attract readers’ attention. To see what are the new connotation that are formed in the present advertisements, they will be compared with advertisements in the older generation. This research is using qualitative descriptive method and the data is coming from the wordings that are printed in those advertisements. The data will be described and analyzed with Discourse of Advertising Theory by Guy Cook and Anchoring Connotation Theory by Roland Barthes, and then the result will be conclude. Through Guy’s theory, this research will reveal all the elements that are used by the producers to build the beauty concept. Through Barthes’ theory, this research will show how the advertisers create new connotation. The result of analysis shows that present beauty concept about ideal body is a beautiful body that are radiated from the outer side of the women’s body and good looking. Beautiful face is a face that looks as young as twenty years old girls. Moreover, the result also shows that producers use current social representative where consumers these days like new, instant products that give maximum result. The result also shows consumers’ doubt over unknown products and women’s fear on fat body and aging.

  13. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  14. 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

  15. Modular 3-D solid finite element model for fatigue analyses of a PWR coolant system

    International Nuclear Information System (INIS)

    Garrido, Oriol Costa; Cizelj, Leon; Simonovski, Igor

    2012-01-01

    Highlights: ► A 3-D model of a reactor coolant system for fatigue usage assessment. ► The performed simulations are a heat transfer and stress analyses. ► The main results are the expected ranges of fatigue loadings. - Abstract: The extension of operational licenses of second generation pressurized water reactor (PWR) nuclear power plants depends to a large extent on the analyses of fatigue usage of the reactor coolant pressure boundary. The reliable estimation of the fatigue usage requires detailed thermal and stress analyses of the affected components. Analyses, based upon the in-service transient loads should be compared to the loads analyzed at the design stage. The thermal and stress transients can be efficiently analyzed using the finite element method. This requires that a 3-D solid model of a given system is discretized with finite elements (FE). The FE mesh density is crucial for both the accuracy and the cost of the analysis. The main goal of the paper is to propose a set of computational tools which assist a user in a deployment of modular spatial FE model of main components of a typical reactor coolant system, e.g., pipes, pressure vessels and pumps. The modularity ensures that the components can be analyzed individually or in a system. Also, individual components can be meshed with different mesh densities, as required by the specifics of the particular transient studied. For optimal accuracy, all components are meshed with hexahedral elements with quadratic interpolation. The performance of the model is demonstrated with simulations performed with a complete two-loop PWR coolant system (RCS). Heat transfer analysis and stress analysis for a complete loading and unloading cycle of the RCS are performed. The main results include expected ranges of fatigue loading for the pipe lines and coolant pump components under the given conditions.

  16. 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

  17. Radioactive corrosion products in circuit of fast reactor loop with dissociating coolant

    International Nuclear Information System (INIS)

    Dolgov, V.M.; Katanaev, A.O.

    1982-01-01

    The results of experimental investigation into depositions of radionuclides of corrosion origin on the surfaces of a reactor-in-pile loop facility with a dissociating coolant are presented. It is stated that the ratio of radionuclides in fixed depositions linearly decreases with decrease of the coolant temperature at the core-condenser section. The element composition of non-fixed compositions quantitatively and qualitatively differs from the composition of structural material, and it is more vividly displayed for the core-condenser section. The main mechanism of circuit contamination with radioactive corrosion products is substantiated: material corrosion in the zones of coolant phase transfer, their remove by the coolant in the core, deposition, activation and wash-out by the coolant from the core surfaces

  18. 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

  19. Effects of LWR coolant environments on fatigue design curves of carbon and low-alloy steels

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1998-03-01

    The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components. Figures I-9.1 through I-9.6 of Appendix I to Section III of the code specify fatigue design curves for structural materials. While effects of reactor coolant environments are not explicitly addressed by the design curves, test data indicate that the Code fatigue curves may not always be adequate in coolant environments. This report summarizes work performed by Argonne National Laboratory on fatigue of carbon and low-alloy steels in light water reactor (LWR) environments. The existing fatigue S-N data have been evaluated to establish the effects of various material and loading variables such as steel type, dissolved oxygen level, strain range, strain rate, temperature, orientation, and sulfur content on the fatigue life of these steels. Statistical models have been developed for estimating the fatigue S-N curves as a function of material, loading, and environmental variables. The results have been used to estimate the probability of fatigue cracking of reactor components. The different methods for incorporating the effects of LWR coolant environments on the ASME Code fatigue design curves are presented

  20. 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

  1. The challenge of modeling fuel–coolant interaction: Part I – Premixing

    Energy Technology Data Exchange (ETDEWEB)

    Meignen, Renaud, E-mail: renaud.meignen@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, IRSN/PSN-RES/SAG, BP 3, 13115 Saint-Paul-Lez-Durance Cedex (France); Picchi, Stephane; Lamome, Julien [Communication and Systèmes, 22 avenue Galilée, 92350 Le Plessis Robinson (France); Raverdy, Bruno [IRSN/PSN-RES/SAG, BP3, 92362 Fontenay aux Roses Cedex (France); Escobar, Sebastian Castrillon [Institut de Radioprotection et de Sûreté Nucléaire, IRSN/PSN-RES/SAG, BP 3, 13115 Saint-Paul-Lez-Durance Cedex (France); Nicaise, Gregory [IRSN/PSN-RES/SAG, BP3, 92362 Fontenay aux Roses Cedex (France)

    2014-12-15

    Highlights: • We present the status modeling of the fuel–coolant interaction premixing stage in the computer code MC3D. • We also propose a general state of the art, highlighting recent improvements in understanding and modeling, remaining difficulties, controversies and needs. • We highlight the need for improving the understanding of the melt fragmentation and oxidation. • The verification basis is presented. - Abstract: Fuel–coolant interaction is a complex mixing process that can occur during the course of a severe accident in a nuclear power plant involving core melting and relocation. Under certain circumstances, a steam explosion might develop during the mixing of the melt and the water and induce a loss of integrity of the containment. Even in the absence of an explosion, studying the mixing phenomenon is also of high interest due to its strong impact on the progression of the accident (debris bed formation, hydrogen production). This article is the first of two aiming at presenting both a status of research and understanding of fuel–coolant interaction and the main characteristics of the model developed in the 3-dimensional computer code MC3D. It is devoted to the premixing phase whereas the second is related to the explosion phase. A special attention is given to major difficulties, uncertainties and needs for further improvements in knowledge and modeling. We discuss more particularly the major phenomena that are melt fragmentation and film boiling heat transfer and the challenges related to modeling melt solidification and oxidation. Some highlights related to the code verification are finally given.

  2. Premixing of corium into water during a Fuel-Coolant Interaction. The models used in the 3 field version of the MC3D code and two examples of validation on Billeau and FARO experiments

    Energy Technology Data Exchange (ETDEWEB)

    Berthoud, G.; Crecy, F. de; Duplat, F.; Meignen, R.; Valette, M. [CEA/Grenoble, DRN/DTP, 17 Avenue des Martyrs, 38054 Grenoble Cedex 9 (France)

    1998-01-01

    This paper presents the <> application of the multiphasic 3D computer code MC3D. This application is devoted to the premixing phase of a Fuel Coolant Interaction (FCI) when large amounts of molten corium flow into water and interact with it. A description of the new features of the model is given (a more complete description of the full model is given in annex). Calculations of Billeau experiments (cold or hot spheres dropped into water) and of a FARO test (<> corium dropped into 5 MPa saturated water) are presented. (author)

  3. 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

  4. Examination of a failed reactor coolant pump rotating assembly from Crystal River Unit 3

    International Nuclear Information System (INIS)

    Hayner, G.O.; Lubnow, T.; Clary, M.

    1990-01-01

    On January 18, 1989, the A reactor coolant pump rotating assembly at the Crystal River Unit 3 Nuclear Power Plant failed during operation. A rotating assembly from this pump had previously failed in 1986. The reactor coolant pump was fabricated by Byron Jackson Pump Division of Borg-Warner Ind. Products, Inc. from UNS S66286 superalloy (Alloy A286). A root cause failure analysis examination was performed on the pump shaft and other components. The failure analysis included shaft vibrational mode and stress analyses, pump clearance and alignment analyses, and detailed destructive examination of the shaft and hydrostatic bearing assemblies. Based on the detailed physical examination of the shaft it was concluded that cracks initiated in the pump shaft at two sites approximately 180 0 apart in a band of shallow, thermally induced fatigue cracks. The cracks initiated at the bottom edge of the motor end shrink fit pad under the shrink fit sleeve supporting the hydrostatic bearing journal. The band of thermally induced fatigue cracks was apparently caused by mixing of cold seal injection water and hot reactor coolant in gaps between the pump shaft and sleeve. The motor end shrink fit was apparently not effective in preventing introduction of the seal injection water to this area. Initial crack propagation occurred by fatigue due to lateral vibration; however, the majority of crack propagation occurred by abnormal torsional fatigue loading induced by contact and sticking between the rotating and stationary portions of the hydrostatic bearing. Final fracture of the shaft occurred by torsional overload. Metallurgical characteristics and mechanical properties of the shaft were within design specification and probably did not significantly influence the cracking process

  5. 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.

  6. 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

  7. Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Joo S.; Diamond, David

    2016-12-06

    A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

  8. Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

    1994-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

  9. Recent bibliography on analytical and sampling problems of a PWR primary coolant Suppl. 4

    International Nuclear Information System (INIS)

    Illy, H.

    1986-09-01

    The 4th supplement of a bibliographical series comprising the analytical and sampling problems of the primary coolant of PWR type reactors covers the literature from 1985 up to July 1986 (220 items). References are listed according to the following topics: boric acid; chloride, chlorine; general; hydrogen isotopes; iodine; iodide; noble gases; oxygen; other elements; radiation monitoring; reactor safety; sampling; water chemistry. (V.N.)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. Modelling transient energy release from molten fuel coolant interaction debris

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-05-01

    A simple model of transient energy release in a Molten Fuel Coolant Interaction is presented. A distributed heat transfer model is used to examine the effect of heat transfer coefficient, time available for rapid energy heat transfer and particle size on transient energy release. The debris is assumed to have an Upper Limit Lognormal distribution. Model predictions are compared with results from the SUW series of experiments which used thermite-generated uranium dioxide molybdenum melts released below the surface of a pool of water. Uncertainties in the physical principles involved in the calculation of energy transfer rates are discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  13. 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)

  14. Analysis on Radioactive Waste Transmutation in Light Water cooled Hyb-WT

    International Nuclear Information System (INIS)

    Hong, Seonghee; Kim, Myung Hyun

    2014-01-01

    A feasibility of realization is much higher in FFHR compared with pure fusion. A combination of plasma fusion source for neutrons with a subcritical reactor at the blanket side has much higher capability in transmutation of waste as well as reactor safety compared with fission reactor options. Fusion-Fission Hybrid Reactor (FFHR) uses various coolants depending on the purpose. It is important that coolant being used should be suitable to reactor purpose, because reactor performance and the design constraints may change depending on the coolant. There are basically two major groups of coolants for FFHR. One group of coolant does not contain Li. They are Na, Pb-Bi, H 2 O and D 2 O. The other group contains Li for tritium breeding. They are Li, LiPb, LiSN, FLIBE and FLiNaBe. Currently, the issue in FFHR is its implication for radioactive waste transmutation (FFHR for WT). Because radioactive wastes of spent nuclear fuel (SNF) are transmuted using fusion neutron source. Therefore a suitable coolant should be used for effective waste transmutation. . In FFHR for WT, LiPb coolant is being used mainly because of tritium production in Li and high neutron economic through reaction in Pb. However different coolants use such as Na, Pb-Bi are used in fast reactors and accelerator driven systems (ADS) having same purpose. In this study, radioactive waste transmutation performance of various coolants mentioned above will be compared and analyzed. Through this study, the coolants are judged primarily for their support to waste transmutation disregarding their limitation to reactor design and tritium breeding capability. First, performance of the light water coolant regarding radioactive waste transmutation was analyzed among various coolants mentioned above. In this paper, performance of radioactive waste transmutation can be known depending on different volume fractions (54.53, 60.27, 97.94vol.%) of the light water. Light water dose required fusion power lower than LiPb due to

  15. 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

  16. Experimental study on utilization of air-borne jet sound in coolant leak detector

    International Nuclear Information System (INIS)

    Hayamizu, Y.; Kitahara, T.; Hayashi, T.; Nishimura, M.

    1975-10-01

    Studies have been undertaken to develop a new coolant leak detection method by the use of a microphone to pick up jet sound generated when pressurized high temperature water is discharged from a pressure boundary into the atmosphere. Leakage was simulated in three shapes, such as two machine-made circular holes and longitudinal and transverse slits in an inlet tube of a blowdown test facility. The measured power level of the jet sound was in agreement with theoretical values calculated from Lighthill's equation. In the study of utilization, this new method has been confirmed as applicable, and to be calculated theoretically for design on 'signal to noise ratio' evaluation. Detection of a small coolant leakage of 1 kg/sec is possible in a recirculation pump room which has large background noise from the pump if a suitable isolation wall, such as hot boxes, is installed between the monitored pipes and the pump. (auth.)

  17. 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

  18. 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)

  19. Predicted effect of power uprating on the water chemistry of commercial boiling water reactors

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Chu, Charles F.; Chang Ching

    2009-01-01

    The approach of power uprating has been adopted by operators of light water reactors in the past few decades in order to increase the power generation efficiency of nuclear reactors. The power uprate strategy is apparently applicable to the three nuclear reactors in Taiwan as well. When choosing among the three types of power uprating, measurement uncertainty, stretch power uprating, and extended power uprating, a deliberate and thorough evaluation is required before a final decision and an optimal selection can be made. One practical way of increasing the reactor power is to deliberately adjust the fuel loading pattern and the control rod pattern and thus to avoid replacing the primary coolant pump with a new one of larger capacity. The power density of the reactor will increase with increasing power, but the mass flow rate in the primary coolant circuit (PCC) of a light water reactor will slightly increase (usually by less than 5 %) or even remain unchanged. Accordingly, an uprated power would induce higher neutron and gamma photon dose rates in the reactor coolant but have a minor or no effect on the mass flow rate of the primary coolant. The radiolysis product concentrations and the electrochemical corrosion potential (ECP) values differ largely in the PCC of a boiling water reactor (BWR). It is very difficult to measure the water chemistry data directly at various locations of an actual reactor. Thus the impact of power uprating on the water chemistry of a BWR operating under hydrogen water chemistry (HWC) can only be theoretically evaluated through computer modelling. In this study, the DEMACE computer code was modified to investigate the impact of power uprating on the water chemistry under a fixed mass flow rate in the primary coolant circuit of a BWR/6 type plant. Simulations were carried out for hydrogen concentrations in feedwater ranging from 0.0 to 2.0 mg . kg -1 and for power levels ranging from 100 % to 120 %. The responses of water chemistry and ECP

  20. 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

  1. 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.

  2. 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.

  3. 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

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

    International Nuclear Information System (INIS)

    Chapman, A.G.

    1987-03-01

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

  5. 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

  6. 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

  7. THE PROBLEM OF ENERGY EFFICIENCY OF THE GEOTHERMAL CIRCULATION SYSTEM IN DIFFERENT MODES OF REINJECTION OF THE COOLANT

    Directory of Open Access Journals (Sweden)

    D. K. Djavatov

    2017-01-01

    Full Text Available Aim. Advanced technologies are crucial for widespread use of geothermal energy to ensure its competitiveness with conventional forms of energy. To date, the basis for the development of geothermal energy is the technology of extracting the heat transfer fluids from the subsoil. There are the following ways to extract the coolant: freeflow; pumping and circular methods. Of greatest interest is the technology to harness the geothermal energy based on geothermal circulatory system (GCS. There is the problem of the right choice of technological parameters for geothermal systems to ensure their effective functioning.Methods. We consider the development of geothermal energy technology based on geothermal circulatory system, as this technology solves the dumping of the waste water containing environmentally harmful substances. In addition to the environmental issues, this technology makes it possible to intensify the process of production and the degree of extraction of thermal resources, which significantly increases the potential for geothermal heat resources in terms of the fuel and energy balance.Findings. Were carried out optimization calculations for Ternairsky deposits of thermal waters. In the calculations, was taken into account the temperature dependence of important characteristics, such as the density and heat capacity of the coolant.Conclusions. There is the critical temperature of the coolant injected, depending on the flow rate and the diameter of the well, ensuring the effective functioning of the geothermal circulatory systems. 

  8. 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

  9. Reactor primary coolant system pipe rupture study. Progress report No. 33, January--June 1975

    International Nuclear Information System (INIS)

    1975-10-01

    The pipe rupture study is designed to extend the understanding of failure-causing mechanisms and to provide improved capability for evaluating reactor piping systems to minimize the probability of failures. Following a detailed review to determine the effort most needed to improve nuclear system piping (Phase 1), analytical and experimental efforts (Phase 2) were started in 1965. This progress report summarizes the recent accomplishments of a broad program in (a) basic fatigue crack growth rate studies focused on LWR primary piping materials in a simulated BWR primary coolant environment, (b) at-reactor tests of the effect of primary coolant environment on the fatigue behavior of piping steels, (c) studies directed at quantifying weld sensitization in Type 304 stainless steel, (d) support studies to characterize the electrochemical potential behavior of a typical BWR primary water environment and (e) special tests related to simulation of fracture surfaces characteristic of IGSCC field failures

  10. 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.)

  11. 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

  12. SIMMER-III applications to fuel-coolant interactions

    Energy Technology Data Exchange (ETDEWEB)

    Morita, K.; Kondo, Sa.; Tobita, Y.; Brear, D.J. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-01-01

    The main purpose of the SIMMER-III code is to provide a numerical simulation of complex multiphase, multicomponent flow problems essential to investigate core disruptive accidents in liquid-metal fast reactors (LMFRs). However, the code is designed to be sufficiently flexible to be applied to a variety of multiphase flows, in addition to LMFR safety issues. In the present study, some typical experiments relating to fuel-coolant interactions (FCIs) have been analyzed by SIMMER-III to demonstrate that the code is applicable to such complex and highly transient multiphase flow situations. It is shown that SIMMER-III can reproduce the premixing phase both in water and sodium systems as well as the propagation of steam explosion. It is thus demonstrated the code is basically capable of simulating integral multiphase thermal-hydraulic problems included in FCI experiments. (author)

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

    International Nuclear Information System (INIS)

    Siddiqui, M.S.

    1992-06-01

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

  14. 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

  15. 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)

  16. 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.)

  17. Reactor coolant pump seal response to loss of cooling

    International Nuclear Information System (INIS)

    Graham, T.; Metcalfe, R.; Burchett, P.

    2000-01-01

    This paper describes the results of a test done to determine the performance of a reactor coolant pump seal for a water cooled nuclear reactor under loss of all cooling conditions. Under these conditions, seal faces can lose their liquid lubricating film and elastomers can rapidly degrade. Temperatures in the seal-cartridge tester reached 230 o C in three hours, at which time the tester was stopped and the temperature increased to 265 o C for a further five hours before cooling was restored. Seal leakage was 'normal' throughout the test. Parts sustained minor damage with no effect on seal integrity. Plant operators were shown to have ample margin beyond their 15 minute allowable reaction time. (author)

  18. 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

  19. 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

  20. 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