WorldWideScience

Sample records for water reactors phwrs

  1. Design of condenser for 500 MWe pressurised heavy water reactors (PHWRs) - a case study

    International Nuclear Information System (INIS)

    Agarwal, N.K.; Subbarao, A.; Chaudhary, K.

    1996-01-01

    Condenser forms the major heat sink in the power plants. In recent years, power plant availability and performance have become great concern to the industry. The detailed design of the condenser and its associated cooling water (CW) system require careful optimisation of parameters which include material selection, cooling water flow rate, condenser surface areas, turbine exhaust pressures etc. This is required to produce a design offering maximum efficiency and reliability and minimum maintenance. The various parameters involved in condenser design are discussed. 5 refs., 1 fig

  2. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    Rastogi, B.P.

    1989-01-01

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  3. Management of radioactive effluents from research Reactors and PHWRs

    International Nuclear Information System (INIS)

    Bodke, S.B.; Surender Kumar; Sinha, P.K.; Budhwar, R.K.; Raj, Kanwar

    2006-01-01

    Indian nuclear power programme is mainly based on pressurized heavy water reactors (PHWRs). In addition we have research reactors namely Apsara, CIRUS, Dhruva at Trombay. The operation and maintenance activities of these reactors generate radioactive liquid waste. These wastes require effective management so that the release of radioactivity to the environment is well within the authorized limits. India is self reliant in the design, erection, commissioning and operation of effluent management system for nuclear reactors. Segregation at source based on nature of effluents and radioactivity content is the first and foremost step in the over all management of liquid effluents. The effluents from the power reactors contain mainly activation products like 3 H. It also contains fission products like 137 Cs. Containment of these radionuclide along with 60 Co, 90 Sr, 131 I plays an important part in liquid waste management. Treatment processes for decontamination of these radionuclide include chemical treatment, ion exchange, evaporation etc. Effluents after treatment are monitored and discharged to the nearby water body after filtration and dilution. The concentrates from the processes are conditioned in cement matrix and disposed in Near Surface Disposal Facilities (NSDFs) co-located at each site. Some times large quantity of effluents with higher radioactivity concentration may get generated from the abnormal operation such as failure of heat exchangers. These effluents are handled on a campaign basis for which adequate storage capacity is provided. The treatment is given taking into consideration the required decontamination factor (DF), capacities of available treatment process, discharge limits and the availability of the dilution water. Similarly large quantities of effluents may get generated during fuel clad failure incident in reactors. In such situation, as in CIRUS large volume of effluent containing higher radioactivity are generated and are managed by delay

  4. Operational experience in water chemistry of PHWRs

    International Nuclear Information System (INIS)

    Krishna Rao, K.S.

    2000-01-01

    The chemistry related problems encountered in the moderator, primary heat transport systems, chemical control in the steam generators and the experience gained in the decontamination campaigns carried out in the primary heat transport systems of Indian PHWRs are highlighted in this paper. (author)

  5. Reactor protection systems of 500 MWe PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Mallik, G; Kelkar, M G; Apte, Ravindra [C and I Group, Nuclear Power Corporation, Mumbai (India)

    1997-03-01

    The 500 MWe PHWR has two totally independent, diverse, fast acting shutdown system called Shutdown System 1 (SDS 1) and Shutdown System 2 (SDS 2). The trip generation circuitry of SDS 1 and SDS 2 are known as Reactor Protection System 1 (RPS 1) and Reactor Protection System 2 (RPS 2) respectively. Some of the features specific to 500 MWe reactors are Core Over Power Protection System (COPPS) based upon in core Self Powered Neutron Detector (SPND) signals, use of local two out of three coincidence logic and adoption of overlap testing for RPS 2, use of Fine Impulse Testing (FIT) in RPS 2, testing of the final control elements namely electro-magnetic clutch of individual Shutoff Rods (SRs) of SDS 1 and all the fast acting valves of SDS 2, etc. The two shutdown systems have totally separate sets of sensors and associated signal processing circuitry as well as physical arrangements. A separate computerised test and monitoring unit is used for each of the two shutdown systems. Use of Programmable Digital Comparator (PDC) unit exclusively for reactor protection systems, has been adopted. The capability of PDC unit is enhanced and communication links are provided for its integration in over all system. The paper describes the design features of reactor protection systems. (author). 3 refs., 7 figs., 3 tabs.

  6. Ageing of coolant channels in nuclear reactors (PHWRs)

    International Nuclear Information System (INIS)

    Mitra, T.L.; Chowdhury, M.K.; Gupta, R.K.; Pandarinathan, P.R.; Seth, V.K.

    1994-01-01

    In PHWRs, ageing of various components takes place due to factors like fast neutron flux, temperature, stress, environment etc. In coolant channel, the most severely affected component due to ageing is pressure tube, though other components like end fitting, calandria tube, garter spring spacer also show ageing to a limited extent. Ageing effects in pressure tube are seen in the form of diametral and axial creep, corrosion, delayed hydrogen cracking and irradiation hardening. In calandria tube and garter spring spacer, creep and hardening are seen though these are not of concern in PHWRs. In end fitting, irradiation embrittlement and abrasion of sealing faces are the areas of concern. Ageing process in these components are the areas of concern. Ageing process in these components are effectively retarded by taking measures like selection of proper material, manufacturing process, control of environmental chemistry, and design modifications. Experience and information gained in various Indian and foreign reactors have been used to improve upon the design in 220 MWe reactors and have formed the basis of design for 500 MWe reactors. (author). 3 refs., 5 figs

  7. An assessment of post-LOCA radiolytic generation of hydrogen in reactor containment of Indian PHWRs

    International Nuclear Information System (INIS)

    Bose, H.; Shah, G.C.; Dutta, S.

    2002-01-01

    Full text: An event-wise assessment has been carried out for the 220 MWe Indian PHWRs of standardized design, to estimate the post-LOCA release of radiolytic hydrogen inside reactor containment, in absence of steam-zirconium reaction. The assessment is based on (i) the dissolved hydrogen concentration build-up in water corresponding to the decaying gamma dose profile and (ii) the rate of concentration dependent mass-transfer of hydrogen from water to gas-space. It is observed that the total radiolytic hydrogen released is about three times less than that obtained by the conventional method of calculation which assumes the radiolytic yield of hydrogen to be equal to the primary yield G(H 2 ) = 0.44 molecules per 100 eV. It is also seen that a major part (∼90 %) of the total release is due to the spillage of fission product irradiated suppression pool water flowing through the core, followed by moderator and suppression pool surface releases respectively

  8. Safety and licensing issues for Indian PHWRs

    International Nuclear Information System (INIS)

    Srinivasan, G.R.; Das, M.

    1997-01-01

    India has achieved competency in design, construction, commissioning and operation of Pressurized Heavy Water Reactor based Nuclear Power Plants and has completed more than 120 reactor operating years with an extremely satisfactory safety record. In this paper, the safety management in NPCIL and operational safety aspects are discussed, licensing and regulatory approach is described and some of the main safety issues for Indian PHWRs are brought out. (author)

  9. The concept of power correction techniques and its use in the reactor regulation and protection systems in Indian PHWRs

    International Nuclear Information System (INIS)

    Vaswani, P.D.; Kelkar, M.G.; Ghoshal, B.; Ashok Kumar, B.

    2010-01-01

    Reactor Power Measurement is an essential part of the Reactor Power Control Loop in PHWRs. None of the available power measuring sensor offers characteristics which allow their direct use in the Reactor Power Control Loop. Thermal power, which is considered as relatively accurate, suffers from measurement delays and is used only as reference. Neutronic power sensors like Ion Chambers and Self Powered Neutron Detectors (SPNDs) which sense instantaneous power suffer from inaccuracies. A technique is required which makes use of both types-reference power and instantaneous power to extract real power information from the signals. This paper describes techniques to calibrate (correct) neutronic power that with the thermal reference power signals. The paper also brings out limitation of the calibration technique. (author)

  10. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  11. Manufacture of fuel and fuel channels and their performance in Indian PHWRs'

    International Nuclear Information System (INIS)

    Kalidas, R.

    2005-01-01

    Nuclear Fuel Complex (NFC) at Hyderabad is conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried our several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

  12. Manufacture of fuel and fuel channels and their performance in Indian PHWRS - an overview

    International Nuclear Information System (INIS)

    Kalidas, R.

    2005-01-01

    Nuclear Fuel Complex (NFC) at Hyderabad is a conglomeration of chemical, metallurgical and mechanical plants, processing uranium and zirconium in two separate streams and culminating in the fuel assembly plant. Apart from manufacturing fuel for Pressurised Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs), NFC is also engaged in the manufacture of reactor core structurals for these reactors. NFC has carried out several technological developments over the years and implemented them for the manufacture of fuel, calandria tubes and pressure tubes for PHWRs. Keeping in pace with the Nuclear Power Programme envisaged by the Department of Atomic Energy, NFC had augmented its production capacities in all these areas. The paper highlights several actions initiated in the areas of fuel design, fuel manufacturing, manufacturing of zirconium alloy core structurals, fuel clad tubes and components and their performance in Indian PHWRs. (author)

  13. Surface analytical and electrochemical characterization of oxide films formed on Incoloy-800 and carbon steel in simulated secondary water chemistry conditions of PHWRs

    International Nuclear Information System (INIS)

    Rangarajan, S.; Sinu, C.; Balaji, V.; Narasimhan, S.V.

    2010-01-01

    The water chemistry in the Steam Generator (SG) Circuits of Indian Pressurized Heavy Water Reactors (PHWRs) is controlled by the all volatile treatment (AVT) procedure, wherein volatile amines are used to maintain the alkaline pH required for minimizing the corrosion of the structural materials. Earlier, Monel and morpholine were used as the Steam Generator material and the alkalizing agent respectively. However, currently they are replaced by Incoloy-800 and Ethanolamine (ETA). ETA was chosen because of its beneficial effects due to low pK b and K d values, loading behaviour on condensate polishing unit (CPU) and also on cost comparison with other amines. Since we have Incoloy-800 on the tube side and Carbon steel(CS) on the shell side in the SG circuits, efforts were taken to study the nature of the oxide films formed on these surfaces and to evaluate the corrosion resistance and electrochemical properties of the same, under simulated secondary water chemistry conditions of PHWRs containing different dissolved oxygen (DO) concentration. In this context, experiments were carried out by exposing finely polished CS and Incoloy -800 coupons to ETA based medium in the presence and absence of Hydrazine (pH: 9.2) at 240 o C under two different DO conditions (< 10 ppb and 200 ppb) for 24 hours. Oxide films formed under these conditions were characterized using SEM, Raman spectroscopy, electrochemical impedance, polarization and Mott-Schottky techniques. Further, studies at a controlled DO level ( < 10 ppb) were carried out for different time durations viz., 7- and 30- days. The composition, surface morphology, oxide thickness, resistance, type of semi-conductivity and defect density of the oxide films were evaluated and correlated with the DO levels and discussed elaborately in this paper. (author)

  14. Development of methods to control radiation field and corrosion in PHWRS

    International Nuclear Information System (INIS)

    Velmurugan, S.

    2015-01-01

    Pressurized Heavy Water Reactors (PHWRs) is the mainstay of Indian Nuclear Power Program. There are 18 PHWRs (220 MWe and 540 MWe) in operation and 4 X 700 MWe PHWRs are under construction. In these reactors, as far as radiation field is concerned, the philosophy of ALARA (As Low As Reasonably Achievable) is followed. The primary coolant system chemistry control is given due consideration during operation so that corrosion of structural material is minimized which in turn controls the radiation field. Development and application of full system Dilute Chemical Decontamination (DCD) process helped to reduce the radiation field in MAPS-1 and 2, RAPS-1 and 2, NAPS-1 and 2 and KAPS-1. PHWR being a tube type reactor, it enables application of full system decontamination to its heavy water primary coolant system. Significant reduction in radiation field and consequent savings in MANREM could be achieved. Attempts are being made to understand the problem created by the release of antimony activities ( 122 Sb and 124 Sb) during chemical decontamination and during planned shutdown. Passivation as a method to control the radiation field and corrosion is being studied. Magnesium ion as a passivator to the ferrite filmed structural materials of PHWRs is being investigated. In addition, as PHWRs uses carbon steel as structural material, the use of passivation as a method to control flow accelerated corrosion (FAC) is also being studied. Magnesium ion gets incorporated in the ferrite film formed over carbon steel structural material and is expected to reduce the solubility of magnetite film thereby the FAC of feeders in PHWRs. (author)

  15. Fast reactor programme in India

    Indian Academy of Sciences (India)

    2015-09-04

    , mainly pressurized heavy water reactors (PHWRs) to .... plug housing 12 absorber rod drive mechanisms is supported on ... state-of-art erection equipments and construction methodologies and .... This decision is taken after.

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

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1991-01-01

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

  17. Water cooled reactor technology: Safety research abstracts no. 1

    International Nuclear Information System (INIS)

    1990-01-01

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

  18. Development of in-situ laser based cutting technique for shock absorber rear nut in pressurized heavy water reactors. CP-2.1

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Jain, R.K.; Upadhyaya, B.N.; Choubey, Ambar; Agrawal, D.K.; Oak, S.M.

    2007-01-01

    We have developed a laser based cutting technique for shock absorber rear nuts in pressurized heavy water reactors (PHWRs). This technique has been successfully used for in-situ laser cutting at RAPS-3 reactor. The technique consists of a motorized compact fixture, which holds a fiber optic beam delivery cutting nozzle and can be operated remotely

  19. In-core fuel management benchmarks for PHWRs

    International Nuclear Information System (INIS)

    1996-06-01

    Under its in-core fuel management activities, the IAEA set up two co-ordinated research programmes (CRPs) on complete in-core fuel management code packages. At a consultant meeting in November 1988 the outline of the CRP on in-core fuel management benchmars for PHWRs was prepared, three benchmarks were specified and the corresponding parameters were defined. At the first research co-ordination meeting in December 1990, seven more benchmarks were specified. The objective of this TECDOC is to provide reference cases for the verification of code packages used for reactor physics and fuel management of PHWRs. 91 refs, figs, tabs

  20. An analysis of workers' tritium concentration in urine samples as a function of time after intake at Korean pressurised heavy water reactors.

    Science.gov (United States)

    Kim, Hee Geun; Kong, Tae Young

    2012-12-01

    In general, internal exposure from tritium at pressurised heavy water reactors (PHWRs) accounts for ∼20-40 % of the total radiation dose. Tritium usually reaches the equilibrium concentration after a few hours inside the body and is then excreted from the body with an effective half-life in the order of 10 d. In this study, tritium metabolism was reviewed using its excretion rate in urine samples of workers at Korean PHWRs. The tritium concentration in workers' urine samples was also measured as a function of time after intake. On the basis of the monitoring results, changes in the tritium concentration inside the body were then analysed.

  1. Post Fukushima safety enhancements in Indian PHWRS

    International Nuclear Information System (INIS)

    Ramasomayajulu, M.; Khot, Pankaj; Chauhan, Ashok

    2016-01-01

    Fukushima event was reviewed in Nuclear Power Corporation of India (NPCIL) and based on these reviews, safety enhancements were identified for Indian PHWRs. Safety enhancements such as additional emergency power sources, enhanced onsite water inventories, external water injection arrangements (Hook up points), measures related to hydrogen management, containment venting provision, seismic trip, mobile pumps, onsite emergency support Centre. These safety enhancements were reviewed by the regulatory body (Atomic Energy Regulatory Board, AERB) and were approved for implementation. Most of these are either implemented or in the advance stage of implementation. The paper elaborates above safety enhancements implemented post Fukushima accident; and preparedness to use these provisions. (author)

  2. Design of a thorium fuelled Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Krishnani, P.D.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Joginder Kumar; Nema, M.K.

    2000-01-01

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

  4. Lessons for PHWRs learned from the Chernobyl accident

    International Nuclear Information System (INIS)

    Waddington, J.G.; Molloy, T.J.

    1996-04-01

    The Atomic Energy Control Board of Canada examined its criteria for licensing nuclear power plants following the accident to the Chernobyl reactor in 1986. The causes of the accident were studied to ascertain whether they revealed any deficiencies in the safety of CANDU PHWRs. A report published in 1987 contained nine recommendations, and this paper revisits these to indicate how they were dealt with by plant owners and the regulatory authority. (author)

  5. Lessons for PHWRs learned from the Chernobyl accident

    International Nuclear Information System (INIS)

    Waddington, J.G.; Molloy, T.J.

    1996-01-01

    The Atomic Energy Control Board of Canada examined its criteria for licensing nuclear power plants following the accident to the Chernobyl reactor in 1986. The causes of the accident were studied to ascertain whether they revealed any deficiencies in the safety of CANDU PHWRs. A report published in 1987 contained nine recommendations, and this paper revisits these to indicate how they were dealt with the plant owners and the regulatory authority

  6. Light water reactor safety

    CERN Document Server

    Pershagen, B

    2013-01-01

    This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

  7. Studies on antimony absorption on Carbon steel (CS) and magnetite coated CS at high temperature to investigate the problem of out of core Sb activity in PHWRs

    International Nuclear Information System (INIS)

    Keny, S.J.; Gokhale, B.K.; Kumbhar, A.G.; Bera, Santanu; Velmurugan, S.

    2014-01-01

    Sb from PHT (primary heat transfer) pump bearings of PHWRs (Pressurized Heavy Water Reactors) goes to the reactor core and gets activated to 121 Sb and 123 Sb. Subsequently, it deposits on out of core surface resulting in radiation exposure to station personnel's apparent high decontamination factors. Sb, thus deposited can't be impassivated by normal decontamination process. Earlier studies indicates lattice substitution of Sb +3 for Fe +2 in magnetite at low doping levels (≤5%). This process, at reactor conditions is yet to be well understood. To formulate an adequate decontamination formulation and methodology and for having insight at Sb deposition mechanism under rector conditions studies are performed

  8. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    2008-06-01

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

  9. Modelling of activity transport in primary heat transport (PHT) system of Indian PHWRs

    International Nuclear Information System (INIS)

    Markandeya, S.G.; Pujari, P.K.; Gandhi, H.C.; Venkateswaran, G.; Narasimhan, S.V.; Krishnarao, K.S.; Mathur, P.K.; Venkat Raj, V.

    2000-01-01

    Nuclear Power plants (NPPs) are designed and built with the aim of minimising the occupational exposure to the operational and maintenance staff. Despite the use of prudently selected materials of construction with high corrosion resistance and adopting very stringent water chemistry controls during operation the build-up of activity in the Primary Heat Transport (PHT) systems of NPPs has been found to be unavoidable. The Indian Pressurised Heavy Water Reactors (PHWRs) are no exception to this. To enable advance planning of maintenance work and the decontamination schedules, it is necessary to perform the off-site calculations to predict the activity buildup in the PHT circuits of the NPPs. A computer code ANUCRUD is under development for predicting the corrosion product and activity transport behaviour in the PHT circuits of Indian PHWRs. The present paper briefly describes some of the salient features of the code ANUCRUD. As a first attempt, preliminary calculations for predicting corrosion product crud concentration buildup in the PHT circuit of the 220 MWe Indian PHWR have been carried out using the code. The findings of these studies are discussed in the paper. Finally, the further improvements proposed to be carried out in the code are also brought out in the paper. (author)

  10. Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

    Energy Technology Data Exchange (ETDEWEB)

    Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

    1994-12-31

    The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  12. Super critical water reactors

    International Nuclear Information System (INIS)

    Dumaz, P.; Antoni, O; Arnoux, P.; Bergeron, A; Renault, C.; Rimpault, G.

    2005-01-01

    Water is used as a calori-porter and moderator in the most major nuclear centers which are actually in function. In the pressurized water reactor (PWR) and boiling water reactor (BWR), water is maintained under critical point of water (21 bar, 374 Centigrade) which limits the efficiency of thermodynamic cycle of energy conversion (yield gain of about 33%) Crossing the critical point, one can then use s upercritical water , the obtained pressure and temperature allow a significant yield gains. In addition, the supercritical water offers important properties. Particularly there is no more possible coexistence between vapor and liquid. Therefore, we don't have more boiling problem, one of the phenomena which limits the specific power of PWR and BWR. Since 1950s, the reactor of supercritical water was the subject of studies more or less detailed but neglected. From the early 1990s, this type of conception benefits of some additional interests. Therefore, in the international term G eneration IV , the supercritical water reactors had been considered as one of the big options for study as Generation IV reactors. In the CEA, an active city has engaged from 1930 with the participation to a European program: The HPWR (High Performance Light Water Reactor). In this contest, the R and D studies are focused on the fields of neutrons, thermodynamic and materials. The CEA intends to pursue a limited effort of R and D in this field, in the framework of international cooperation, preferring the study of versions of rapid spectrum. (author)

  13. Supercritical Water Reactors

    International Nuclear Information System (INIS)

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

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  14. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

    The operation of a pressurized water reactor (PWR) is described with respect to the procedure for a unit start-up. The systems details and numerical data are for a four loop PWR station of the design proposed for Sizewell-'B', United Kingdom. A description is given of: the initial conditions, filling the reactor coolant system (RCS), heat-up and pressurisation of the RCS, secondary system preparations, reactor start-up, and reactivity control at power. (UK)

  15. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  16. Enriched uranium cycles in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Mazzola, A.

    1994-01-01

    A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

  17. Post irradiation examination of garter springs from Indian PHWRs

    International Nuclear Information System (INIS)

    Dubey, J.S.; Shah, Priti Kotak; Mishra, Prerna; Singh, H.N.; Alur, V.D.; Kumar, Ashwini; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

    2013-12-01

    Irradiated Zr-2.5Nb-0.5Cu garter springs, belonging to Indian Pressurised Heavy Water Reactors, which had experienced 8 to 10 Effective Full Power Years of operation were subjected to visual, dimensional, chemical, metallographic examination and relevant mechanical tests. Methodology of the tests conducted and results are presented. The digital photographs were used to measure the inner and outer circumferences by image processing. The hydrogen (H) content in the spring coils were measured using Differential Scanning Calorimetry (DSC). In the stretch test, all the irradiated GSs were found to require an additional load, as compared to unirradiated GS, to produce a given amount of residual extension which indicated that the irradiated GSs had undergone significant irradiation hardening. The crush test results showed that the minimum load required to crush the coil or cause a sudden sideways shift in the grips was higher than 400 N/coil, much higher than the design load. The test results indicated that the irradiated GS, after 10 EFPY of operation, have adequate strength and ductility to continue to meet the design intent. Mechanical tests were carried out on irradiated girdle wires taken out of the loose fit garter springs (GS) from (NAPS-1, ∼ 8.5 EFPY) and tight fit garter spring from KAPS-2 (∼ 8.0 EFPY) PHWRs. Tensile tests on the irradiated girdle wires, showed irradiation hardening in the material and reduction in ductility. The irradiated girdle wires have around 4 to 5% residual ductility level against the 15% ductility of unirradiated wire. The fracture surfaces of the irradiated as well as the un-irradiated girdle wires were observed in SEM. (author)

  18. Reactor water chemistry control

    International Nuclear Information System (INIS)

    Kundu, A.K.

    2010-01-01

    Tarapur Atomic Power Station - 1 and 2 (TAPS) is a twin unit Boiling Water Reactors (BWRs) built in 1960's and operating presently at 160MWe. TAPS -1 and 2 are one of the vintage reactors operating in the world and belongs to earlier generation of BWRs has completed 40 years of successful, commercial and safe operation. In 1980s, both the reactors were de-rated from 660MWth to 530MWth due to leaks in the Secondary Steam Generators (SSGs). In BWR the feed water acts as the primary coolant which dissipates the fission heat and thermalises the fast neutrons generated in the core due to nuclear fission reaction and under goes boiling in the Reactor Pressure Vessel (RPV) to produce steam. Under the high reactor temperature and pressure, RPV and the primary system materials are highly susceptible to corrosion. In order to avoid local concentration of the chemicals in the RPV of BWR, chemical additives are not recommended for corrosion prevention of the system materials. So to prevent corrosion of the RPV and the primary system materials, corrosion resistant materials like stainless steel (of grade SS304, SS304L and SS316LN) is used as the structural material for most of the primary system components. In case of feed water system, main pipe lines are of carbon steel and the heater shell materials are of carbon steel lined with SS whereas the feed water heater tubes are of SS-304. In addition to the choice of materials, another equally important factor for corrosion prevention and corrosion mitigation of the system materials is maintaining highly pure water quality and strict water chemistry regime for both the feed water and the primary coolant, during operation and shutdown of the reactor. This also helps in controlled migration of corrosion product to and from the reactor core and to reduce radiation field build up across the primary system materials. Experience in this field over four decades added to the incorporation of modern techniques in detection of low

  19. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  1. WATER BOILER REACTOR

    Science.gov (United States)

    King, L.D.P.

    1960-11-22

    As its name implies, this reactor utilizes an aqueous solution of a fissionable element salt, and is also conventional in that it contains a heat exchanger cooling coil immersed in the fuel. Its novelty lies in the utilization of a cylindrical reactor vessel to provide a critical region having a large and constant interface with a supernatant vapor region, and the use of a hollow sleeve coolant member suspended from the cover assembly in coaxial relation with the reactor vessel. Cool water is circulated inside this hollow coolant member, and a gap between its outer wall and the reactor vessel is used to carry off radiolytic gases for recombination in an external catalyst chamber. The central passage of the coolant member defines a reflux condenser passage into which the externally recombined gases are returned and condensed. The large and constant interface between fuel solution and vapor region prevents the formation of large bubbles and minimizes the amount of fuel salt carried off by water vapor, thus making possible higher flux densities, specific powers and power densities.

  2. Containment design for Indian PHWRs - evolution and future trends

    International Nuclear Information System (INIS)

    Chatterjee, S.K.; Srinivasan, G.R.; Das, M.; Prakash, P.; Mulgund, S.

    1994-01-01

    The design of containment systems for PHWRs in India has undergone progressive improvements to enhance their reliability and effectiveness. The state-of-the-art containment design incorporates a double containment structure for minimizing radioactivity release to the environment, a completely passive vapour suppression system with huge suppression pool for limiting pressure build-up during postulated LOCA and various engineered systems for depressurizing the containment and cleaning the containment environment following an accident. The containment related Engineered Safety Features (ESFS) include Reactor Building (RB) coolers, Primary Containment Controlled Discharge (PCCD) system, Primary Containment Filtration and Pump-Back (PCFPB) system and Secondary Containment Filtered Recirculation and Purge (SCFRP) system. Studies indicate that the unique feature of double containment with huge suppression pool at basement and associated ESFs not only ensures near zero ground level release during Design Basis Accident (DBA) conditions, but also provides adequate assurance for containment integrity even in beyond DBA scenario. In this paper, an outline of the containment design evolution in Indian PHWRs is presented and salient features of standardized containment design are highlighted. Important containment related studies are discussed and outstanding safety issues viz. hydrogen generation and management, containment venting, containment over pressure capability, etc. are addressed. (author). 16 refs., 1 tab., 8 figs

  3. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  4. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  5. Evaluation of advanced hot conditioning process for PHWRS

    International Nuclear Information System (INIS)

    Chandramohan, P.; Srinivasan, M.P.; Velmurugan, S.

    2015-01-01

    Hot-conditioning/hot functional test process is carried out to the PHT system of reactor before reactor going to critical/operational. The process is aimed in checking the component functionalities at high temperature and high pressure conditions, the process also checks/removes the suspended corrosion products in heat transport circuit. This process leads to formation of a passive or corrosion oxide film on the heat transport circuit surfaces which protects/mitigates the corrosion of the system circuits during the operation of plant. Major concerned alloy in the Primary Heat Transport (PHT) system of Indian PHWRs during the hot conditioning process and also during operation is the carbon steel due to its high corrosion. Hot-conditioning process mitigates the corrosion of carbon steel by the formation of iron oxide (Fe 3 O 4 ) as major oxide phase layer on the carbon steel surface with a typical thickness of 1.0 μm with particle size of 1μm after 336 h of process at 250 °C. But this passive oxide film thickness increase with time of operation of system with c.a. 10μm for 2.2 EFYP. The protectiveness of passive layer can be further enhanced by reducing the particle sizes in the passive film to nano meter range. The process can impact on the compactness of passive oxide layer with reduced pores in the oxide layer and properties of the nano nature oxide (transport properties) impacting the corrosion mitigation. The corrosion mitigation reduce the source term in the activated corrosion product generation. To achieve this a new process 'Advanced hot conditioning' was developed in water steam chemistry division, BARC for getting a passive oxide film with a lowered particle size in the passive film. The AHC process with 1g/L of PEG-8000 at 250 °C for 336 h showed a particle size <100 nm. The process was tested under the normal operating conditions as function of the time, the corrosion parameter like oxide film thickness, corrosion rate and metal ion

  6. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  7. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  8. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

    A steam turbine and an electric generator are connected by way of a speed convertor. The speed convertor is controlled so that the number of rotation of the electric generator is constant irrespective of the speed change of the steam turbine. A shaft coupler is disposed between the turbine and the electric generator or between the turbine and a water injection pump. With such a constitution, the steam turbine and the electric generator are connected by way of the speed convertor, and since the number of revolution of the electric generator is controlled to be constant, the change of the number of rotation of the turbine can be controlled irrespective of the change of the number of rotation of the electric generator. Accordingly, the flow rate of the injection water from the water injection pump to a reactor pressure vessel can be controlled freely thereby enabling to supply stable electric power. (T.M.)

  9. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  10. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay

    International Nuclear Information System (INIS)

    Dhami, P.S; Yadav, J.S; Agarwal, K.

    2017-01-01

    Exploitation of the abundant thorium resources to meet sustained energy demand forms the basis of the Indian nuclear energy programme. To gain reprocessing experience in thorium fuel cycle, thoria was irradiated in research reactor CIRUS in early sixties. Later in eighties, thoria bundles were used for initial flux flattening in some of the pressurized heavy water reactors (PHWRs). The research reactor irradiated thoria contained small content (∼ 2-3ppm) of "2"3"2U in "2"3"3U product, which did not pose any significant radiological problems during processing in Uranium Thorium Separation Facility (UTSF), Trombay. Thoria irradiated in PHWRs on discharge contained (∼ 0.5-1.5% "2"3"3U with significant "2"3"2U content (100-500 ppm) requiring special radiological attention. Based on the experience from UTSF, a new facility viz. Power Reactor Thoria Reprocessing Facility (PRTRF), Trombay was built which was hot commissioned in the year 2015

  11. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  12. Development of expanded type plugging technique for leaky tubes of steam generators of Indian PHWRs

    International Nuclear Information System (INIS)

    Das, Nirupam; Samuel, K.A.; Joemon, V.; Rupani, B.B.

    2006-01-01

    Steam generators are very important component of Nuclear Power Plant (NPP), as they are part of Primary Heat Transport (PHT) system of Pressurised Heavy Water Reactors (PHWRs). A nuclear power plant of 220 MWe capacity has four mushroom type steam generators, each consisting of 1830 U-tubes (16 mm outside diameter and 1 mm wall thickness) made of Incoloy-800 material. The tubes of 'tube and shell type steam generator' act as the pressure boundary of PHT System. Any structural failure of these tubes may lead to release of radioactivity along with plant outage and significant economic loss. Hence, it is necessary to plug the leaky tubes for continued and safe operation of a steam generator. An expanded type plugging technique has been developed at Reactor Engineering Division to plug the leaky tubes. This plugging technique is selected because of low residual stress imparted in the adjacent 'tube to tube-sheet' joints. This plug meets the various codal requirements of steam generator. A number of qualification trials have been carried out with such plugs in the mock up facility. The expanded plugs meet the design requirements for pull out strength and leak-tightness. This paper describes the design concept of the plug, developmental aspects and qualification of the plugging technique. (author)

  13. 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.)

  14. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

    The principles of nuclear, thermal and hydraulic performance calculations for water cooled reactors are discussed. The principles are illustrated by describing their implementation in the UKAEA PATRIARCH scheme of computer codes. This material was originally delivered as a course of lectures at the Technical University of Helsinki in Summer of 1969.

  15. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  16. Indigeneous design and development of differential pressure reducing valves for PHWRs (Paper No. 055)

    International Nuclear Information System (INIS)

    Soni, N.L.; Agrawal, R.C.; Chandra, Rajesh

    1987-02-01

    On load fuelling of Pressurised Heavy Water Reactors (PHWRs) is being achieved with the help of Fuelling Machine (F/M). Various actuations are to be carried out inside the F/M magazine pressure housing with the help of high pressure water hydraulic actuators. A constant differential pressure is required to be maintained between pressurized magazine housing and the actuators-supply line for proper operation of the actuators which are to be located inside it. This is achieved with the help of the Differential Pressure Reducing Valve (DPRV). So far these valves have been procured only from a single foreign supplier. In March 1980, the price of each valve was US dollars 3100.00. Dependence on a single foreign supplier may create problems of timely procurement. An effort was made to design and manufacture the DPRV indigensouly meeting the stringent specifications. Theoretical study of single acting DPRV was carried out and design criteria were established. The valve was tested for its performance and was found satisfactory. (author). 8 figs

  17. A new method for detecting pressure tube failures in Indian PHWRs

    International Nuclear Information System (INIS)

    Sharma, V.K.; Gupta, V.K.

    2000-01-01

    For the annulus gas system (AGS) of the standardised Indian pressurised heavy water reactor, an elaborate pressure tube (PT) crack monitoring and detection system is envisaged to ensure safety through leak-before-break. The parameters that are monitored relate to the detection of D 2 O moisture leaking in from the primary heat transport (PHT) system through a cracked PT. Since a slow build-up of moisture in the AGS may also occur for reasons other than PT failure, it is desirable that a diverse measurement technique should be available. This paper suggests such a technique, based on the observation that a small reference concentration of fission gases is normally present in the annulus gas. This concentration would change sharply upon PT failure, when the heavy water from the leaking PHT system releases the dissolved fission gas content into the annulus. This paper presents a theoretical study of the parameters that influence the build-up of fission product noble gases in the AGS and shows that leakage rates as low as 10 g h -1 from a PT crack can be detected in a few tens of minutes by this method. This is expected to substantially increase the available time between the leak detection and the PT failure, thus serving as an important tool in meeting the leak-before-break criterion of a critical component in PHWRs. (orig.)

  18. Construction management of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bohra, S.A.; Sharma, P.D.

    2006-01-01

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  19. Next generation light water reactors

    International Nuclear Information System (INIS)

    Omoto, Akira

    1992-01-01

    In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

  20. Advanced light-water reactors

    International Nuclear Information System (INIS)

    Golay, M.W.; Todreas, N.E.

    1990-01-01

    Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

  1. Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Dhruvanarayana, L.; Gupta, H.; Bharathkumar, M.

    1996-01-01

    Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D 2 0 hydraulics, H 2 0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of valves

  2. Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

    1984-01-01

    The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

  3. Chemistry in water reactors

    International Nuclear Information System (INIS)

    Hermansson, H.P.; Norring, K.

    1994-01-01

    The international conference Chemistry in Water Reactors was arranged in Nice 24-27/04/1994 by the French Nuclear Energy Society. Examples of technical program areas were primary chemistry, operational experience, fundamental studies and new technology. Furthermore there were sessions about radiation field build-up, hydrogen chemistry, electro-chemistry, condensate polishing, decontamination and chemical cleaning. The conference gave the impression that there are some areas that are going to be more important than others during the next few years to come. Cladding integrity: Professor Ishigure from Japan emphasized that cladding integrity is a subject of great concern, especially with respect to waterside corrosion, deposition and release of crud. Chemistry control: The control of the iron/nickel concentration quotient seems to be not as important as previously considered. The future operation of a nuclear power plant is going to require a better control of the water chemistry than achievable today. One example of this is solubility control via regulation in BWR. Trends in USA: means an increasing use of hydrogen, minimization of SCC/IASCC, minimization of radiation fields by thorough chemistry control, guarding fuel integrity by minimization of cladding corrosion and minimization of flow assisted corrosion. Stellite replacement: The search for replacement materials will continue. Secondary side crevice chemistry: Modeling and practical studies are required to increase knowledge about the crevice chemistry and how it develops under plant operation conditions. Inhibitors: Inhibitors for IGSCC and IGA as well for the primary- (zinc) as for the secondary side (Ti) should be studied. The effects and mode of operation of the inhibitors should be documented. Chemical cleaning: of heat transfer surfaces will be an important subject. Prophylactic cleaning at regular intervals could be one mode of operation

  4. The heavy water reactors

    International Nuclear Information System (INIS)

    Brudermueller, G.

    1976-01-01

    This is a survey of the development so far of this reactor line which is in operation all over the world in various types (e.g. BHWR, PHWR). MZFR and the CANDU-type reactors are discussed in more detail. (UA) [de

  5. Nuclear reactor in deep water

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    Events during October 1980, when the Indian Point 2 nuclear reactor was flooded by almost 500 000 litres of water from the Hudson river, are traced and the jumble of human errors and equipment failures chronicled. Possible damage which could result from the reactor getting wet and from thermal shock are considered. (U.K.)

  6. Experiences in the application of quality control and quality assurance programmes in water reactor fuel fabrication

    International Nuclear Information System (INIS)

    Balaramamoorthy, K.; Vijayaraghavan, R.; Kulkarni, P.G.; Raghavan, S.V.; Bandyopadhyay, A.K.

    1984-01-01

    Nuclear fuel for Research Reactors and Pressurised Heavy Water Reactors (PHWRs) is being fabricated in India for a period of over two decades. The fuel is produced to conform to stringent quality control specifications. Generally, the performance of the fuel has been very good in the reactors. This is not only due to the high quality workmanship in the various stages of production but also to the meticulous care exercised in the planning and application of quality control and quality assurance procedures. For the nuclear fuel used in Water Reactors, extensive material specifications have been compiled and they are periodically reviewed and revised. The specifications cover various aspects such as metallurgical and mechanical properties, non-destructive testing, dimensional and visual standard requirements. Similarly, detailed manufacturing engineering instructions (MEIs) and quality control instructions (QCIs) have been drawn. For any deviations from the specified requirements, design concession committee considers all deviations and acceptance or rejection criteria are evolved. In this task, the design concession committee is supported by experimentation in various laboratories of the Department of Atomic Energy. The Quality Assurance procedures have been evolved over a long period of time. They generally conform to the latest code and recommended guides of IAEA regarding Quality Assurance in the manufacture of fuel. (orig.)

  7. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  8. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

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

  9. Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Water simulation of sodium reactors

    International Nuclear Information System (INIS)

    Grewal, S.S.; Gluekler, E.L.

    1981-01-01

    The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system

  11. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    Drevon, G.

    1983-01-01

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

  12. Heat exchangers in heavy water reactor systems

    International Nuclear Information System (INIS)

    Mehta, S.K.

    1988-01-01

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

  13. Optimisation of parameters of DCD for PHWRs

    International Nuclear Information System (INIS)

    Velmurugan, S.; Sathyaseelan, V.S.; Narasimhan, S.V.; Mathur, P.K.

    1991-01-01

    Decontamination formulation based on EDTA, Oxalic acid, Citric acid was evaluated for its efficacy in removing oxide layers of PHWR. An ion exchange system which was specifically suitable for fission product dominated contamination in PHWRs was optimised for the reagent regeneration stage of the decontamination process. An analysis of the nature of the complexed metal species formed in the dissolution process and Electrochemical measurements were employed as a tool to follow the course of oxide removal during the dissolution process. An attempt was made to understand the redeposition behaviour of various isotopes during the decontamination process. SEM and ESCA studies of metal coupons before and after the dissolution process were used to analyse the deposits in the above context. The pick up of DCD reagents on the ion exchangers and material compatibility tests on Carbon steel, Monel-400 and Zircaloy-2 with the decontaminant under the conditions of decontamination experiment are reported. (author)

  14. Thermal-hydraulic R and D infrastructure for water cooled reactors of the Indian nuclear power program

    International Nuclear Information System (INIS)

    Vijayan, P.K.; Jain, V.; Saha, D.; Sinha, R.K.

    2009-01-01

    R and D has been the critical ingredient of Indian Nuclear Power Program from the very inception. Approach to R and D infrastructure has been closely associated with the three-stage nuclear power program that was crafted on the basis of available resources and technology in the short-term and energy security in the long-term. Early R and D efforts were directed at technologies relevant to Pressurized Heavy Water Reactors (PHWRs) which are currently the mainstay of Indian nuclear power program. Lately, the R and D program has been steered towards the design and development of advanced and innovative reactors with the twin objective of utilization of abundant thorium and to meet the future challenges to nuclear power such as enhanced safety and reliability, better economy, proliferation resistance etc. Advanced Heavy Water Reactor (AHWR) is an Indian innovative reactor currently being developed to realize the above objectives. Extensive R and D infrastructure has been created to validate the system design and various passive concepts being incorporated in the AHWR. This paper provides a brief review of R and D infrastructure that has been developed at Bhabha Atomic Research Centre for thermal-hydraulic investigations for water-cooled reactors of Indian nuclear power program. (author)

  15. Thorium in heavy water reactors

    International Nuclear Information System (INIS)

    Andersson, G.

    1984-12-01

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

  16. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  17. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

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

  19. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

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

  20. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  1. The safety of light water reactors

    International Nuclear Information System (INIS)

    Pershagen, B.

    1986-04-01

    The book describes the principles and practices of reactor safety as applied to the design, regulation and operation of both pressurized water reactors and boiling water reactors. The central part of the book is devoted to methods and results of safety analysis. Some significant events are described, notably the Three Mile Island accident. The book concludes with a chapter on the PIUS principle of inherent reactor safety as applied to the SECURE type of reactor developed in Sweden. (G.B.)

  2. Ageing management of Indian PHWRs - safety aspects

    International Nuclear Information System (INIS)

    Kapoor, R.K.; Sah, B.M.L.; Das, M.; Srinivasan, G.R.

    1994-01-01

    Ageing management has now become a vital area of concern. Ageing management includes determination of degradation factors, taking various steps to determine present conditions of systems, structures and components and taking mitigating steps. It also includes updating, modernization, refurbishment etc. It is important that ageing management starts right from the time of commissioning of the unit and is treated as a continuous process, and a parallel effort to the normal running of the plant. Thus elaborate research and development efforts are required to be instituted. Life extension could have a high benefit to cost ratio. Various steps to ensure safety in ageing management are listed. Selection of critical items, condition monitoring and life estimation of the same and a chronological check sheet from 0 to 60 years, for Indian PHWRs is explained. Areas where future research and development and other efforts need to be directed is listed. The paper concludes emphasizing the need for a systematized approach to ageing management. It recommends intensive research in certain listed areas and suggests standing committees in specialized areas to tap Indian experience in other industries and establishments. A safety guide is also required to be produced to cover all facets of ageing management. (author). 3 appendices

  3. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

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

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

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

  5. 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

  6. Evolution of the design of fuel handling control system in 220 MWe Indian PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Dhruvanarayana, L; Gupta, H; Bharathkumar, M [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    Following two CANDU type reactors at Rajasthan (RAPS-1 and 2), three nuclear power stations, each of two units of 220 MWe has been in operation at Rajasthan (RAPS-1 and 2). Madras (MAPS-1 and 2). Narora (NAPS-1 and 2) and Kakrapar (KAPS-1 and 2). Two more stations, also of 220 MWe capacity, are under construction at Rajasthan (RAPP-3 and 4) and Kaiga (Kaiga-1 and 2). These are natural uranium fuelled pressurized heavy water cooled and heavy water moderated reactors (PHWRs). The two units at Rajasthan viz RAPS-1 and 2, were built with the technical collaboration with Canada, and the rest of the units have been designed and built indigenously, incorporating a number of modifications, particularly in the on-power refuelling system. The evolution of the design of the Fuel Handling Control systems of these reactors, taking into consideration operational needs, safety aspects and maintainability are highlighted in this paper. A combination of hydraulic and electronic control has been provided to enable the operations. In RAPS-1 and 2, hardwired electronic controls were provided, while in MAPS-1 and 2, the hardwired system was improved. From NAPS onwards, a computerized control system with hardwired interlock logic has been provided. New devices like coarse-fine potentiometers, special oil filled potentiometer assembly, rectilinear potentiometers etc., were specified from NAPS onwards. Positioning logic is computerized providing flexibility and expendability. Digital panel meters and indicating lamps have been provided for manual mode operations, while CRT (cathode-ray tube) monitors help in computer mode operations. Hydraulic controls which comprise D{sub 2}0 hydraulics, H{sub 2}0 hydraulics and oil hydraulics have been improved from NAPS onwards. Hydraulic panels have been relocated in accessible areas to reduce radiation doses and for better maintainability. All electric drives including X and Y drives were modified as hydraulic drives for better control. New types of

  7. Advances in heavy water reactors

    International Nuclear Information System (INIS)

    1994-03-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The Technical Committee Meeting (TCM) on Advances in Heavy Water Reactors was organized by the IAEA in the framework of the activities of the International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) and hosted by the Atomic Energy of Canada Limited. Sixty-five participants from nine countries (Canada, Czech Republic, India, German, Japan, Republic of Korea, Pakistan, Romania and USA) and the IAEA attended the TCM. Thirty-four papers were presented and discussed in five sessions. A separate abstract was prepared for each of these papers. All recommendations which were addressed by the participants of the Technical Committee meeting to the IWGATWR have been submitted to the 5th IWGATWR meeting in September 1993. They were reviewed and used as input for the preparation of the IAEA programme in the area of advanced water cooled reactors. This TCM was mainly oriented towards advances in HWRs and on projects which are now in the design process and under discussion. Refs, figs and tabs

  8. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    Science.gov (United States)

    Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

    2008-12-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

  9. Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

    2008-01-01

    Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

  10. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  11. Feed water control device in a reactor

    International Nuclear Information System (INIS)

    Okutani, Tetsuro.

    1984-01-01

    Purpose: To prevent substantial fluctuations of the water level in a nuclear reactor and always keep a constant standard level under any operation condition. Constitution: When the causes for fluctuating the reactor water level is resulted, a certain amount of correction signal is added to a level deviation signal for the difference between the reactor standard level and the actual reactor water level to control the flow rate of the feed water pump depending on the addition signal. If reactor scram should occur, for instance, a level correction signal changing stepwise depending on a scram signal is outputted and added to the level deviation signal. As the result, the flow rate of feed water sent into the reactor just after the scram is increased, whereby the lowering in the reactor water level upon scram can be decreased as compared with the case where no such level compensation signal is inputted. (Kamimura, M.)

  12. Water treatment process for nuclear reactors

    International Nuclear Information System (INIS)

    Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

    1992-01-01

    Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigate water treatment process for nuclear reactor utilization. Analysis of output water chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerates to obtain the optimum quantity of pure water which reached to 15 cubic meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30 %. output water chemistry agree with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined.5 fig., 3 tab

  13. Water treatment process for nuclear reactors

    International Nuclear Information System (INIS)

    Marwan, M.A.; Khattab, M.S.; Hanna, A.N.

    1993-01-01

    Water treatment for purification is very important in reactor cooling systems as well as in many industrial applications. Since impurities in water are main source of problems, it is necessary to achieve and maintain high purity of water before utilization in reactor cooling systems. The present work investigates water treatment process for nuclear reactor utilization. Analysis of outwater chemistry proved that demineralizing process is an appropriate method. Extensive experiments were conducted to determine economical concentration of the regenerants to obtain the optimum quantity of pure water which reached to 15 cubic-meter instead of 10 cubic-meter per regeneration. Running cost is consequently decreased by about 30%. Output water chemistry agrees with the recommended specifications for reactor utilization. The radionuclides produced in the primary cooling water due to reactor operation are determined. It is found that 70% of radioactive contaminants are retained by purification through resin of reactor filter. Decontamination factor and filter efficiency are also determined

  14. Reactor water level measuring device

    International Nuclear Information System (INIS)

    Kuroki, Reiji; Asano, Tamotsu.

    1996-01-01

    A condensation vessel is connected to the upper portion of a reactor pressure vessel by way of a pipeline. The lower portion of the condensation vessel is connected to a low pressure side of a differential pressure transmission device by way of a reference leg pipeline. The high pressure side of the differential pressure transmission device is connected to the lower portion of the pressure vessel by way of a pipeline. The condensation vessel is equipped with a temperature sensor. When a temperature of a gas phase portion in the condensation vessel is lowered below a predetermined level, and incondensible gases in the condensation vessel starts to be dissolved in water, signals are sent from the temperature sensor to a control device and a control valve is opened. With such a constitution, CRD driving water flows into the condensation vessel, and water in which gases at the upper portion of the condensation vessel is dissolved flows into the pressure vessel by way of a pipeline. Then, gases dissolved in a reference water column in the reference leg pipeline are eliminated and the value of a reference water pressure does not change even upon abrupt lowering of pressure. (I.N.)

  15. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

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

  16. To the analysis of reactor noise in boiling water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1972-01-01

    The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

  17. High performance light water reactor

    International Nuclear Information System (INIS)

    Squarer, D.; Schulenberg, T.; Struwe, D.; Oka, Y.; Bittermann, D.; Aksan, N.; Maraczy, C.; Kyrki-Rajamaeki, R.; Souyri, A.; Dumaz, P.

    2003-01-01

    The objective of the high performance light water reactor (HPLWR) project is to assess the merit and economic feasibility of a high efficiency LWR operating at thermodynamically supercritical regime. An efficiency of approximately 44% is expected. To accomplish this objective, a highly qualified team of European research institutes and industrial partners together with the University of Tokyo is assessing the major issues pertaining to a new reactor concept, under the co-sponsorship of the European Commission. The assessment has emphasized the recent advancement achieved in this area by Japan. Additionally, it accounts for advanced European reactor design requirements, recent improvements, practical design aspects, availability of plant components and the availability of high temperature materials. The final objective of this project is to reach a conclusion on the potential of the HPLWR to help sustain the nuclear option, by supplying competitively priced electricity, as well as to continue the nuclear competence in LWR technology. The following is a brief summary of the main project achievements:-A state-of-the-art review of supercritical water-cooled reactors has been performed for the HPLWR project.-Extensive studies have been performed in the last 10 years by the University of Tokyo. Therefore, a 'reference design', developed by the University of Tokyo, was selected in order to assess the available technological tools (i.e. computer codes, analyses, advanced materials, water chemistry, etc.). Design data and results of the analysis were supplied by the University of Tokyo. A benchmark problem, based on the 'reference design' was defined for neutronics calculations and several partners of the HPLWR project carried out independent analyses. The results of these analyses, which in addition help to 'calibrate' the codes, have guided the assessment of the core and the design of an improved HPLWR fuel assembly. Preliminary selection was made for the HPLWR scale

  18. Water level monitoring device in nuclear reactor

    International Nuclear Information System (INIS)

    Miura, Kiyohide; Otake, Tomohiro.

    1988-01-01

    Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

  19. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Tanaka, Koji; Egashira, Yasuo; Shimada, Fumie; Igarashi, Noboru.

    1983-01-01

    Purpose: To save a low temperature reactor water clean-up system indispensable so far and significantly simplify the system by carrying out the reactor water clean-up solely in a high temperature reactor water clean-up system. Constitution: The reactor water clean-up device comprises a high temperature clean-up pump and a high temperature adsorption device for inorganic adsorbents. The high temperature adsorption device is filled with amphoteric ion adsorbing inorganic adsorbents, or amphoteric ion adsorbing inorganic adsorbents and anionic adsorbing inorganic adsorbents. The reactor water clean-up device introduces reactor water by the high temperature clean-up pump through a recycling system to the high temperature adsorption device for inorganic adsorbents. Since cations such as cobalt ions and anions such as chlorine ions in the reactor water are simultaneously removed in the device, a low temperature reactor water clean-up system which has been indispensable so far can be saved to realize the significant simplification for the entire system. (Seki, T.)

  20. Advances in light water reactor technologies

    CERN Document Server

    Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

    2010-01-01

    ""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

  1. TA-2 Water Boiler Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Durbin, M.E.; Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m 3 of low-level solid radioactive waste and 35 m 3 of mixed waste. 15 refs., 25 figs., 3 tabs

  2. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  3. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  4. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  5. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

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

  6. Heavy water moderated tubular type nuclear reactor

    International Nuclear Information System (INIS)

    Oohashi, Masahisa.

    1986-01-01

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

  7. Light water reactor safety research project

    International Nuclear Information System (INIS)

    Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

    1980-07-01

    The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

  8. Boiling water reactor fuel bundle

    International Nuclear Information System (INIS)

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

  9. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  10. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

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

  11. Localized corrosion problems in water reactors

    International Nuclear Information System (INIS)

    Coriou, Henri.

    1977-01-01

    Main localized etching on the structure materials of water reactors are studied: stress corrosion on stainless steel 304 (B.W.R), stress corrosion, 'wall thinning' and denting of Inconel 600 vapor generator tubes (P.W.R.). Some mechanisms are examined and practical exemples in reactors are described. Various possible cures are presented [fr

  12. 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

  13. Water feeding method upon reactor isolation

    International Nuclear Information System (INIS)

    Sasaki, Koichi; Takahara, Kuniaki; Hamamura, Kenji; Arakawa, Masahiro.

    1990-01-01

    The present invention concerns a method of feeding water upon reactor isolation in a plural loop type reactor having a plurality of reactor cooling systems. Water can be injected to a plurality of pools even if the pressure between the pools is not balanced and the water level in the reactor cooling system is optimally controlled. That is, water can be injected in accordance with the amount required for each of the pools by setting the opening of a turbine inlet steam control valve to somewhat higher than the cooling system pressure of the highest pressure loop. Water feeding devices upon reactor isolation were required by the same number as that for the reactor cooling systems. Whereas since pumps and turbines are used in common without worsening the water injection controllability to each of the loops according to the method of this invention and, accordingly, the cost performance can be improved. Further, since the opening degree of the turbine inlet steam control valve is controlled while making the difference pressure constant between the turbine inlet pressure and the pump exhaust pressure, the amount of the turbine exhausted steams can be reduced and, further, water injection controllability of the flow rate control valve in the injection line is improved. (I.S.)

  14. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  15. Reactor water quality degradation suppressing method upon reactor start up

    International Nuclear Information System (INIS)

    Maeda, Katsuharu.

    1993-01-01

    Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

  16. Water chemistry in WWER reactors

    International Nuclear Information System (INIS)

    Yurmanov, V.A.; Mamet, V.A.; Shestakov, Yu.M.; Amosov, M.M.

    1997-01-01

    In this paper ''Water Chemistry in WWER Reactors'', are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH T values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ''soft decontamination'' involving changing the KOH concentration and, hence, the pH T before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs

  17. Water chemistry in WWER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yurmanov, V A; Mamet, V A; Shestakov, Yu M; Amosov, M M [All-Russian Scientific Research Inst. for Nuclear Power Plants Operation, Moscow (Russian Federation)

    1997-02-01

    In this paper ``Water Chemistry in WWER Reactors``, are briefly described the 30 WWERs in Russian and the Ukraine, and are pointed out the essential differences between the 440s and 1000s. The primary coolant in the six loops of the former type operates at 270-290 deg. C, while the four loops of the latter type are at 290-320 deg. C. Performance of the fuel has been generally good with some fission product activities emanating from tramp uranium. Incidents causing unusually high fission product levels were overheating of the 16th fuel load at Kola NPP in 1990 by a reduced coolant flow, and fuel defects at Novovoronezh NPP resulting from deposits of carbon and corrosion products. Organic carbon, depositing from the coolant in regions of high turbulence (i.e. at the spacer grids), provokes corrosion product deposition. The source of the organic is not known. New chemistry guidelines have been implemented since 1992-93 for Russian and Ukrainian WWERs. These include higher pH{sub T} values (7.0-7.1 as opposed to 6.6-6.9) and tighter controls on oxygen and impurities. Lower dose rates in steam generator channels are reported. Significant reduction in operator doses are achieved by these methods coupled with a ``soft decontamination`` involving changing the KOH concentration and, hence, the pH{sub T} before shutdown. The benefits of hydrazine treatment for deoxygenating feedwater and coolant prior to start up, for injecting before shutdown and for general chemistry control on radiation fields are described. (author). 7 refs, 9 figs, 8 tabs.

  18. Physics of pressurized water reactors

    International Nuclear Information System (INIS)

    Gruen, A.

    1980-01-01

    The objective of this lecture is to demonstrate typical problems and solutions encountered in the design and operation of PWR power plants. The examples selected for illustration refer to PWR's of KWU design and to results of KWU design methods. In order to understand the physics of a power reactor it is necessary to have some knowledge of the structure and design of the power plant system of which the reactor is a part. It is therefore assumed that the reader is familiar with the design of the more important components and systems of a PWR, such as fuel assemblies, control assemblies, core lay-out, reactor coolant system, instrumentation. (author)

  19. Water-immersion type ship reactor

    International Nuclear Information System (INIS)

    Okada, Hiroki; Yamamura, Toshio.

    1996-01-01

    In a water immersion-type ship reactor in which a water-tight wall is formed around a pressure vessel by way of an air permeable heat insulation layer and immersing the wall under water in a reactor container, a pressure equalizing means equipped with a back flow check valve and introducing a gas in a gas phase portion above the water level of the container into a water tight wall and a relief valve for releasing the gas in the water tight wall to the reactor container are disposed on the water tight wall. When the pressure in the water tight wall exceeds the pressure in the container, the gas in the water tight wall is released from the relief valve to the reactor container. On the contrary, when the pressure in the container exceeds the pressure in the water tight wall, the gas in the gas phase portion is flown from the pressure equalizing means equipped with a back flow check valve to the inside of the water tight wall. Thus, a differential pressure between both of them is kept around 0kg/cm 2 . A large differential pressure is not exerted on the water tight wall thereby capable of preventing rupture of them to improve reliability, as well as the thickness of the plate can be decreased thereby enabling to moderate the design for the pressure resistance and reduce the weight. (N.H.)

  20. Emergency cooling of presurized water reactor

    International Nuclear Information System (INIS)

    Sykora, D.

    1981-01-01

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

  1. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

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

    International Nuclear Information System (INIS)

    1989-07-01

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

  3. Pressurised water reactor in the UK

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Since the Three Mile Island accident there has been much debate about the safety considerations of Pressurised Water Reactors. Their development will continue throughout the world but it will be based upon the lessons learned from that unfortunate accident. In the United Kingdom there is a public enquiry discussing all aspects of the reactor. The papers given in this book provide an informed addition to the literature. The design, safety and licensing and construction of a pressurised water reactor system are discussed in detail. Considerations stemming from the Three Mile Island accident are presented

  4. Hydriding failure in water reactor fuel elements

    International Nuclear Information System (INIS)

    Sah, D.N.; Ramadasan, E.; Unnikrishnan, K.

    1980-01-01

    Hydriding of the zircaloy cladding has been one of the important causes of failure in water reactor fuel elements. This report reviews the causes, the mechanisms and the methods for prevention of hydriding failure in zircaloy clad water reactor fuel elements. The different types of hydriding of zircaloy cladding have been classified. Various factors influencing zircaloy hydriding from internal and external sources in an operating fuel element have been brought out. The findings of post-irradiation examination of fuel elements from Indian reactors, with respect to clad hydriding and features of hydriding failure are included. (author)

  5. Water injection device for reactor container

    International Nuclear Information System (INIS)

    Sakaki, Isao.

    1996-01-01

    A pressure vessel incorporating a reactor core is placed and secured on a pedestal in a dry well of a reactor container. A pedestal water injection line is disposed opened at one end in a pedestal cavity passing through the side wall of the pedestal and led at the other end to the outside of the reactor container. A substitution dry well spray line is connected to a spray header disposed at the upper portion of the dry well. When the pressure vessel should be damaged by a molten reactor core and the molten reactor core should drop to the dry well upon occurrence of an accident, the molten reactor core on the floor of the pedestal is cooled by water injection from the pedestal water injection line. At the same time, the elevation of the pressure and the temperature in the reactor container is suppressed by the water injection of the substitution dry well spray line. This can avoid large scaled release of radioactive materials to the environmental circumference. (I.N.)

  6. Emergency water supply facility for nuclear reactor

    International Nuclear Information System (INIS)

    Karasawa, Toru

    1998-01-01

    Water is stored previously in an equipment storage pit disposed on an operator floor of a reactor building instead of a condensate storage vessel. Upon occurrence of an emergency, water is supplied from the equipment storage pit by way of a sucking pipeline to a pump of a high pressure reactor core water injection circuit and a pump of a reactor-isolation cooling circuit to supply water to a reactor. The equipment storage pit is arranged in a building so that the depth thereof is determined to keep the required amount of water by storing water at a level lower than the lower end of a pool gate during normal operation. Water is also supplied from the equipment storage pit by way of a supply pipeline to a spent fuel storage pool on the operation floor of the reactor building. Namely, water is supplied to the spent fuel storage pool by a pump of a fuel pool cooling and cleaning circuit. This can eliminate a suppression pool cleaning circuit. (I.N.)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  8. Water desalination using different capacity reactors options

    International Nuclear Information System (INIS)

    Alonso, G.; Vargas, S.; Del Valle, E.; Ramirez, R.

    2010-01-01

    The Northwest region of Mexico has a deficit of potable water, along this necessity is the region growth, which requires of additional energy capacity, cogeneration of potable water production and nuclear electricity is an option to be assessed. In this paper we will perform an economical comparison for cogeneration using a big reactor, the AP1000, and a medium size reactor, the IRIS, both of them are PWR type reactors and will be coupled to the desalination plant using the same method. For this cogeneration case we will assess the best reactor option that can cover both needs using the maximum potable water production for two different desalination methods: Multistage Flash Distillation and Multi-effect Distillation. (authors)

  9. New lineup of light water reactors

    International Nuclear Information System (INIS)

    Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

    2007-01-01

    Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

  10. Inherently safe light water reactors

    International Nuclear Information System (INIS)

    Ise, Takeharu

    1987-01-01

    Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

  11. Accident sequences evaluation using SFATs for low power and shutdown operation of pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Kim, Chansoo; Chung, Chang-Hyun; Yang, Huichang

    2004-01-01

    To maintain the level of defense-in-depth safety of Pressurized Heavy Water Reactor (PHWR) during LP/SD operation, the qualitative risk evaluation methods such as Safety Function Assessment Trees (SFATs) are required. Therefore SFATs are suggested to assess and manage the PHWR safety in LP/SD. Before this study, safety functions of PHWR were classified into 7 groups; Reactivity Control, Core Cooling, Secondary Heat Removal, Primary Heat Transport Inventory, Essential Electrical Power, Cooling Water, and Containment Integrity. The SFATs for PHWR LP/SD operations were developed along with the Plant Outage Status (POS) variation, and totally 38 SFATs were developed for Wolsung Unit 2. For the verification of SFATs logics developed, top 5 accident sequences those contribute the CDF of PHWR were selected, and plant safety status were evaluated for those accident sequences. Accident sequences such as DCC-4 (Dual Control Computer Failure), CL4-16 (Total Loss of Class IV Power), and FWPV-11 (Loss of Feedwater Supply to SG due to Failure of Pumps/Values) were included. In this research the evaluation of plant safety status by accident sequences using SFATs and the verification of the SFATs were performed. Through the verification of SFAT logics, the enhancements to the internal logics of the SFATs were made, and the dependencies between safety systems and support systems were considered. It is expected the defense-in-depth evaluation model of PHW just as SFATs can be utilized in the configuration risk management program (CRMP) development and improve technical specifications development for Korean PHWRs. (author)

  12. Method of measuring reactor water level

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1979-01-01

    Purpose: To provide a water level measuring system so that a reactor water level detecting signal can be corrected in correspondence to a recirculation flow, thereby to carry out a correct water level detection in a wide range of the reactor. Method: According to the operation record of a precursor reactor, the ratio Δh of the lowering of the water level due to the recirculation flow is lowered in proportion to the ratiowith respect to the rated differential pressure of the recirculation flow. Accordingly, the flow of recirculation pump is measured by an elbow differential pressure generator utilizing an elbow of a pipe, and the measured value is multiplied by a gain by a ratio setter, and therefter, an addition computation is carried out by an adder for correcting the signal from a water level detector. When the signal from the water level detector is corrected in this manner, the influence of the lowering of the water level due to the recirculation flow can be removed, and an interlocker predetermined in the defined water level can be actuated, thus the influence of the dynamic pressure due to the recirculation flow acting on the instrumental pipe line detecting the reactor water level can be removed effectively. (Yoshino, Y.)

  13. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  14. Requirements for light water reactors

    International Nuclear Information System (INIS)

    Hedin, F.

    2009-01-01

    The EUR (European Utilities Requirements) is an organization founded in 1991 whose aim was to write down the European specifications and requirements for the future reactors of third generation. EUR gathers most of the nuclear power producers of Europe. The EUR document has been built on the large and varied experience of EUR members and can be used to elaborate invitations to tender for nuclear projects. 4000 requirements only for the nuclear part of the plant are listed, among which we have: -) the probability of core meltdown for a reactor must be less than 10 -6 per year, -) the service life of every component that is not replaceable must be 60 years, -) the capacity of the spent fuel pool must be sufficient to store 10-15 years of production without clearing out. The EUR document is both open and complete: every topic has been considered, it does not favor any type of reactor but can ban any technology that is too risky or has an unfavourable feedback experience. The assessment of the conformity with the EUR document of 7 reactor projects (BWR 90/, EPR, EP1000, SWR1000, ABWR, AP1000 and VVER-AES-92) has already be made. (A.C.)

  15. Effect of design improvements on ALARA exposures in PHWRs

    International Nuclear Information System (INIS)

    Nair, S.N.; Mohandas, P.V.; Gupta, Ashok; Hussain, S.A.

    2000-01-01

    Design improvements in Indian PHWRs over last thirty years have remarkably reduced the occupational and public exposures. Some of the radiologically offending systems were altogether changed, equipment was judiciously relocated and at component level reduction in number and improvement in quality was carried out. As a result the collective occupational exposures could be brought down by a factor of about 4-5 and average public exposure by a factor of about 10. Since the design improvements are continuous ongoing processes further reduction in exposure will be definitely brought in the coming years. (author)

  16. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

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

    International Nuclear Information System (INIS)

    1988-10-01

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

  18. Reliability of reactor plant water cleanup pumps

    International Nuclear Information System (INIS)

    Pearson, J.L.

    1979-01-01

    Carolina Power and Light Company's Brunswick 2 nuclear plant experienced a high reactor water cleanup pump-failure rate until inlet temperature and flow were reduced and mechanical modifications were implemented. Failures have been zero for about one year, and water cleanup efficiency has increased

  19. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  20. Utility requirements for advanced light water reactors

    International Nuclear Information System (INIS)

    Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

    1996-01-01

    The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

  1. Safety aspects of pressurised water reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This submission to the Health and Safety Executive has been prepared by the Institution of Professional Civil Servants (IPCS) as a contribution to the debate on safety aspects associated with Pressurized Water Reactors (PWRs). Although supporting an energy policy which includes the development of nuclear power, assurances are sought on a number of safety issues if it is decided that this should be generated by a PWR-type reactor. These issues are listed. In particular the following are mentioned: the wider publication of design information, the use of elastic-plastic fracture mechanics as the basis for determining pressure vessel integrity, the failure rate of steam generating units, water coolant quality control, greater investigation of two-phase flow accident conditions, the components of the reactor cooling system and training of reactor personnel in the understanding of LOCA effects. (U.K.)

  2. 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)

  3. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  4. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  5. Steam explosions in light water reactors

    International Nuclear Information System (INIS)

    1981-01-01

    The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

  6. Evolution of shutdown mechanism for PHWRs

    International Nuclear Information System (INIS)

    Singh, Manjit; Govindarajan, G.

    1997-01-01

    In 500 MWe PHWR, there are two independent fast acting shutdown systems namely (1) mechanical shut-off rod system and (2) liquid poison injection system. Both systems are independently capable of keeping the reactor in sub-critical condition during long shutdown. Mechanical shut-off rod system being primary shutdown system calls for a very high reliability of operation as well as effectiveness, which are mainly governed by its ability to operate within a very short time and the magnitude of negative reactivity worth it can provide. Mechanical shut-off rods are normally parked above the core by shut-off rod drive mechanism. On receiving a scram signal, shut-off rods are released from the holding electromagnetic clutch and fall under gravity into the core. This paper discusses the salient features of mechanical shut-off rod system. A brief account of detailed design and development of sub-assemblies of shut-off rod drive mechanism is also presented. (author)

  7. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Nanis, R.

    2000-01-01

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

  8. Reactor water clean-up device

    International Nuclear Information System (INIS)

    Sawa, Toshio; Takahashi, Sankichi; Takashima, Yoshie.

    1983-01-01

    Purpose: To efficiently eliminate radioactive materials such as iron oxide and cobalt ions with less heat loss by the use of an electrode assembly applied with a direct current. Constitution: In a reactor water clean-up device adapted to pass reactor water through an electrode assembly comprising at least a pair of anode and cathode applied with a direct current to eliminate various types of ions contained in the reactor water by way of the electrolysis or charge neutralization at the anode, the cathode is constituted with a corrosion resistant grid-like or porous metal plate and a layer to the upper portion of the metal plate filled with a plurality of metal spheres of about 1 - 5 mm diameter, and the anode is made of insoluble porous or spirally wound metal material. (Seki, T.)

  9. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-09-01

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

  10. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  11. Instrumentation and control in Indian PHWRs - evolution and vision for future

    International Nuclear Information System (INIS)

    Umesh, Chandra

    2004-01-01

    Full text: Presently there are twelve Pressurized Heavy Water Reactors (PHWRs) under operation and six plants are under construction. The instrumentation and control of these plants has evolved from RAPS-1 onwards and has relied upon indigenous technology development efforts. The evolution process of I and C can be broadly divided into three phases comprising of RAPS and MAPS, NAPS and KAPS and KAIGA and RAPS-3,4. The technologies employed in various areas namely field instrumentation, monitoring and signal processing, interlock logic systems, process control, reactor regulation, protection logic and signal processing, radiation monitoring and operator information systems have evolved from plant to plant mostly independently on need basis -due to additional functional requirements or to avoid obsolescence. A unified approach to C and I was attempted in the 500 MWe design by evolving the concept of DCPIS. However due to tight plant schedules, technologies of third phase were adopted in TAPP-3 and 4 with some changes and additions. In the plants under construction at Kaiga-3 and 4 and RAPP- 5 and 6 also, the I and C technologies are similar to Kaiga-1,2 with incremental changes as required. The technologies presently employed in the PHWR plants have functioned quite satisfactorily and during last year overall plant availability factor was about 90%, which is among the best in the world, The I and C technologies employed in these plants is simple, easy to understand and maintain as well as low cost and hence it is suitable for Indian needs. In older plants, problems sometimes arise due to ageing, obsolescence and lack of documentation in some areas. Some computer-based systems have been affected by noise through power supplies and ground connections necessitating a comprehensive evaluation of these issues. However, there is a need to enhance the level of instrumentation and control from this sound base to achieve lower plant cost and improved safety. This can be

  12. ULTRA SCWR+: Practical advanced water reactor concepts

    International Nuclear Information System (INIS)

    Duffey, Romney; Khartabil, Hussam; Kuran, Sermet; Zhou, Tracy; Pioro, Igor

    2008-01-01

    Modern thermal power plants now utilize supercritical steam cycles with thermal efficiencies of over 45%. Recent developments have lead to Ultra-SuperCritical (USC) systems, which adopt reheat turbines that can attain efficiencies of over 50%. Because these turbines are already developed, demonstrated and deployed worldwide, and use existing and traditional steam cycle technology, the simplest nuclear advance is to utilize these proven thermal cycle conditions by coupling this turbine type to a reactor. This development direction is fundamentally counter to the usual approach of adopting high-temperature gas-cooled (helium-cooled) reactor cycles, for which turbines have yet to be demonstrated on commercial scale unlike the supercritical steam turbines. The ULTRA (Ultra-supercritical Light water Thermal ReActor) SCWR+ concept adopts the fundamental design approach of matching a water and steam-cooled reactor to the ultra-supercritical steam cycle, adopting the existing and planned thermal power plant turbines. The HP and IP sections are fed with conditions of 25 MPa/625degC and 7 MPa/700degC, respectively, to achieve operating plant thermal efficiencies in excess of 50%, with a direct turbine cycle. By using such low-pressure reheated steam, this concept also adopts technology that was explored and used many years ago in existing water reactors, with the potential to produce large quantities of low cost heat, which can be used for other industrial and district processes. Pressure-Tube (PT) reactors are suitable for adoption of this design approach and, in addition, have other advantages that will significantly improve water-cooled reactor technology. These additional advantages include enhanced safety and improved resource utilization and proliferation resistance. This paper describes the PT-SCWR+ concept and its potential enhancements. (author)

  13. Beyond the light water reactor

    International Nuclear Information System (INIS)

    Nero, A.V. Jr.

    1980-01-01

    One of the strong interests in examining alternative nuclear fuel cycles is to search for schemes that are more efficient than LWRs in their use of uranium, but that do not carry the additional proliferation hazard associated with widespread plutonium utilization. One possibility is to improve the uranium efficiency of current reactor types by other means than recycling. A second possibility, offering greater potential for improvement, is to utilize thorium-uranium fuel cycles in which uranium-233 is denatured by the addition of uranium-238, making enrichment necessary to yield weapons-usable material. The bulk of the reactor's fuel material would be thorium-232, so that most of the fissile material produced would be uranium-233. It is important to recognize that these two possibilities - once through improvements and denatured thorium-uranium - could be introduced sequentially in reactor types that are currently in use. Fuel cycles may change over time, but it is equally significant from the point of view of non-proliferation that they will also vary from place to place and, most importantly, from country to country. The author argues that alternative nuclear power systems and a slower growth may affect the diversion of nuclear materials to weapons. A real question, though, is whether we have time to explore the possibilities. It has become apparent that predictions made of the growth rate for nuclear power were too high. The 1000 large power plants the US was to have by the year 2000 have been reduced to fewer than 300. This reduces the pressure, resulting from our limited uranium resources, to push the LMFBR. Extra time gives us a chance to examine the possibilities

  14. EPRI program in water reactor safety

    International Nuclear Information System (INIS)

    Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

    1975-01-01

    The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

  15. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

    This meeting contains articles of the Water Reactor Fuel Performance Meeting 2008 of Korean Nuclear Society, Atomic Energy Society of Japan, Chinese Nuclear Society, European Nuclear Society and American Nuclear Society. It was held on Oct. 19-23, 2008 in Seoul, Korea and subject of Meeting is 'New Clear' Fuel - A green energy solution. This proceedings is comprised of 5 tracks. The main topic titles of track are as follows: Advances in water reactor fuel technology, Fuel performance and operational experience, Transient fuel behavior and safety-related issues, Fuel cycle, spent fuel storage and transportations and Fuel modeling and analysis. (Yi, J. H.)

  16. The water chemistry of CANDU PHW reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-01-01

    This review will discuss the chemistry of the three major water circuits in a CANDU-PHW reactor, viz., the Primary Heat Transport (PHT) water, the moderator and the boiler water. An important consideration for the PHT chemistry is the control of corrosion and of the transport of corrosion products to minimize the growth of radiation fields. In new reactors the PHT will be allowed to boil, requiring reconsideration of the methods used to radiolytic oxygen and elevate the pH. Separation of the moderator from the PHT in the pressure-tubed CANDU design permits better optimization of the chemistry of each system, avoiding the compromises necessary when the same water serves both functions. Major objectives in moderator chemistry are to control (a) the radiolytic decomposition of D 2 0; (b) the concentration of soluble neutron poisons added to adjust reactivity; and (c) the chemistry of shutdown systems. The boiler water and its feed water are treated to avoid boiler tube corrosion, both during normal operation and when perturbations are caused to the feed by, for example, leaks in the condenser tubes which permit ingress of untreated condenser cooling water. Development of a system for automatic analysis and control of feed water to give rapid, reliable response to abnormal conditions is a novel feature which has been developed for incorporation in future CANDU-PHW reactors. (author)

  17. Fabrication of high performance components for Indian nuclear reactors

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2011-01-01

    Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

  18. 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

  19. Evolution of nuclear reactor containments in India: Addressing the present day challenges

    Energy Technology Data Exchange (ETDEWEB)

    Kakodkar, Anil, E-mail: kakodkar@barc.gov.in

    2014-04-01

    Indigenously developed Pressurized Heavy Water Reactors (PHWRs) that form the backbone of current stage of nuclear power development in India have seen continuous evolution of their containment systems. This evolution that has taken place over implementation of 18 PHWRs (200/220/540 MWe) has encompassed all aspects of containment design, viz. the structural system, energy management system, radio-activity management and hydrogen management system. As a part of ongoing efforts toward strengthening of safety performance, India is also ready with the design of Advance Heavy Water Reactor (AHWR), which represents a technology demonstrator for advanced reactor systems and for thorium utilization. This reactor has a number of improved passive safety features and it is capable of meeting the demanding safety challenges that future reactor system would be expected to meet as a result of emerging expectations in the background of accidents over the past three decades viz. those at Three Mile Island (1979), Chernobyl (1986) and most recently at Fukushima (2011). In this lecture I shall focus on the evolution of nuclear reactor containments in India and highlight the design, associated structural and thermal hydraulics safety assessment made over the years for the improvement of containment performance.

  20. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Soppet, W.K.; Shack, W.J.

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  1. Historical perspective of thermal reactor safety in light water reactors

    International Nuclear Information System (INIS)

    Levy, S.

    1986-01-01

    A brief history of thermal reactor safety in U.S. light water reactors is provided in this paper. Important shortcomings in safety philosophy evolution versus time are identified and potential corrective actions are suggested. It should be recognized, that this analysis represents only one person's opinion and that most historical accountings reflect the author's biases and specific areas of knowledge. In that sense, many of the examples used in this paper are related to heat transfer and fluid flow safety issues, which explains why it has been included in a Thermal Hydraulics session. One additional note of caution: the value of hindsight and the selective nature of human memory when looking at the past cannot be overemphasized in any historical perspective

  2. 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.)

  3. Developmental Light-Water Reactor Program

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1989-12-01

    This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

  4. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    Moore, Mark; George, Robert; Spence, Harry; Nguyen, John

    1992-01-01

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  5. Quality assurance plan, Westinghouse Water Reactor Divisions

    Energy Technology Data Exchange (ETDEWEB)

    1976-03-01

    The Quality Assurance Program used by Westinghouse Nuclear Energy Systems Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements.

  6. Light-water reactor accident classification

    International Nuclear Information System (INIS)

    Washburn, B.W.

    1980-02-01

    The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

  7. Facilitation of decommissioning light water reactors

    International Nuclear Information System (INIS)

    Moore, E.B. Jr.

    1979-12-01

    Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented. A wide range of facilitation methods - from improved documentation to special decommissioning tools and techniques - is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given

  8. Hydrogen and water reactor safety: proceedings

    International Nuclear Information System (INIS)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability

  9. Hydrogen and water reactor safety: proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    Separate abstracts were prepared for papers presented in the following areas of interest: 1) hydrogen research programs; 2) hydrogen behavior during light water reactor accidents; 3) combustible gas generation; 4) hydrogen transport and mixing; 5) combustion modeling and experiments; 6) accelerated flames and detonations; 7) combustion mitigation and control; and 8) equipment survivability.

  10. General description of advanced heavy water reactor

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  11. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  12. Light water reactor safeguards system evaluation

    International Nuclear Information System (INIS)

    Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

    1978-01-01

    A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

  13. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

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

  14. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

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

  15. Trends in light water reactor dosimetry programs

    International Nuclear Information System (INIS)

    Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

    1977-01-01

    Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

  16. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1983-01-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

  17. Penn State advanced light water reactor concept

    International Nuclear Information System (INIS)

    Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

    1987-01-01

    The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

  18. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  19. Computer based C and I systems in Indian PHWRs

    International Nuclear Information System (INIS)

    Govindarajan, G.; Sharma, M.P.

    1997-01-01

    Benefits of programmable digital technology have been well recognized and employment of computer based systems in Indian PHWRs has evolved in a phased manner, keeping in view the regulatory requirements for their use. In the initial phase some operator information functions and control of on-power fuel handling system were implemented and then some systems performing control and safety functions have been employed. The availability of powerful microcomputer hardware at reasonable cost and indigenous capability in design and execution has encouraged wider use of digital technology in the nuclear power programme. To achieve the desired level of quality and reliability, the hardware modules for the implementation of these systems in the plants under construction, have been standardized and methodology for software verification and validation has been evolved. A large number of C and I functions including those for equipment diagnostics are being implemented. The paper describes the various applications of computers in Indian NPPs and their current status of implementation. (author)

  20. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  1. Life extension of containment structures of Indian PHWRs

    International Nuclear Information System (INIS)

    Roy, Raghupati; Garg, R.P.; Verma, U.S.P.

    2006-01-01

    Containment structures prevent radioactivity release in the event of any postulated Design Basis Accident (DBA) so that the level of radiation in the external environment is within acceptable limits. Containment structures of Indian PHWRs are typically unlined prestressed concrete structures, which are required to maintain its leak tightness characteristics and strength under DBA during the life of the structure. As nuclear power plant structures age, a number of degradation mechanisms begin to affect critical containment structure. Depending on the type and severity of these degradation mechanisms, its adverse effect on the leak tightness and pressure carrying capacity can be significant. Since the containment structures of Indian PHWRs are unlined, the leak tightness characteristics are solely dependent on the concrete properties and the prestressing material. Prestressing, which is introduced to control the deformation and strength requirement, is affected due to aging. Hence, adequacy of prestressing during the life of the structure to withstand internal pressure and the related leak tightness must be ensured for life extension of prestressed concrete containment structure in view of their significant long term losses. Prevention of corrosion in prestressing steel and assessment of the same at the end of extended design life of the structure, require utmost attention in view of their catastrophic nature of failure. This paper describes the various degradation mechanisms pertaining to concrete and their effect on the leak tightness characteristics and the strength requirement. The issues related to prestressing are also discussed in detail in this paper. The requirement of periodic monitoring of the containment structure for assessing its deformation and leak tightness characteristics and development of database for life extension of containment structure is also addressed in this paper. This paper also discusses the various provisions and measures, which are

  2. Towards intrinsically safe light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  3. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification

  4. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  5. Advanced light water reactor plant

    International Nuclear Information System (INIS)

    Giedraityte, Zivile

    2008-01-01

    For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

  6. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

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

    International Nuclear Information System (INIS)

    Matsushita, T.; Susaki, S.

    1980-01-01

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

  8. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

    During irradiation, nuclear fuel changes volume, primarily through swelling. This swelling is caused by the fission products and in particular by the volatile ones such as krypton and xenon, called fission gas. Fission gas behaviour needs to be reliably predicted in order to make better use of nuclear fuel, a factor which can help to achieve the economic competitiveness required by today's markets. These proceedings communicate the results of an international seminar which reviewed recent progress in the field of fission gas behaviour in light water reactor fuel and sought to improve the models used in computer codes predicting fission gas release. State-of-the-art knowledge is presented for both uranium-oxide and mixed-oxide fuels loaded in water reactors. (author)

  9. The chemistry of water reactor fuel

    International Nuclear Information System (INIS)

    Potter, P.E.

    1990-01-01

    In this paper, the authors discuss features of the changes in chemical constitution which occur in fuel and fuel rods for water reactors during operation and in fault conditions. The fuel for water reactors consists of pellets of urania (UO 2 ) clad in Zircaloy. An essential step in the prediction of the fate of all the radionuclides in a fault or accident is to possess a detailed knowledge of their chemical behavior at all stages of the development of such incidents. In this paper, the authors consider: the chemical constitution of fuel during operation at temperatures corresponding to rather low ratings, together with a quite detailed discussion of the chemistry within the fuel-clad gap; the behavior of fuel subjected to higher temperatures and ratings than those of contemporary fuel; and the changes in constitution on failure of fuel rods in fault or accident conditions

  10. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  11. Materials technologies of light water reactors

    International Nuclear Information System (INIS)

    Begley, R.

    1984-01-01

    Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

  12. Passive systems for light water reactors

    International Nuclear Information System (INIS)

    Adinolfi, R.; Noviello, L.

    1990-01-01

    The paper reviews the most original concepts that have been considered in Italy for the back-fitting of the nuclear power plants in order to reduce the probability and the importance of the release to the environment in case of a core melt. With reference either to BWR or PWR, passive concepts have been considered for back-fitting in the following areas: pump seals damage prevention and ECCS passive operation; reactor passive depressurization; molten reactor core passive cooling; metal containment passive water cooling through a water tank located at high level; containment isolation improvement through a sealing system; containment leaks control and limitation of environmental release. In addition some considerations will be made on the protection against external events introduced from the beginning on the PUN design either on building and equipment lay-out either on structure design. (author). 5 figs

  13. WRAP: a water reactor analysis package

    International Nuclear Information System (INIS)

    Anderson, M.M.

    1977-06-01

    The modular computational system known as the Water Reactor Analysis Package (WRAP) has been developed at the Savannah River Laboratory. WRAP is essentially a reprogrammed version of the RELAP4 computer code with an extensively restructured input format, a dynamic dimensioning capability and additional computational capabilities such as an automatic steady-state option for pressurized water reactors and an automatic restart capability with provision for renodalization. The report describes the capabilities of WRAP at its current stage of development. The addition of new capabilities (e.g., a BWR steady-state capability), the inclusion of improved models (e.g., models in RELAP4/M0D8) and the development of improved numerical techniques to reduce execution time are being planned at this time

  14. Advanced light water reactors for the nineties

    International Nuclear Information System (INIS)

    Ross, F.A.; Sugnet, W.R.

    1987-01-01

    The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

  15. High conversion heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyawaki, Yoshio; Wakabayashi, Toshio.

    1989-01-01

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

  16. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  17. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  18. Light-water reactor safety analysis codes

    International Nuclear Information System (INIS)

    Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

    1980-01-01

    A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

  19. Thermal calculations for water cooled research reactors

    International Nuclear Information System (INIS)

    Fabrega, S.

    1979-01-01

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

  20. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  1. Towards intrinsically safe light-water reactors

    International Nuclear Information System (INIS)

    Hannerz, K.

    1983-02-01

    The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

  2. Reactor core of light water-cooled reactor

    International Nuclear Information System (INIS)

    Miwa, Jun-ichi; Aoyama, Motoo; Mochida, Takaaki.

    1996-01-01

    In a reactor core of a light water cooled reactor, the center of the fuel rods or moderating rods situated at the outermost circumference among control rods or moderating rods are connected to divide a lattice region into an inner fuel region and an outer moderator region. In this case, the area ratio of the moderating region to the fuel region is determined to greater than 0.81 for every cross section of the fuel region. The moderating region at the outer side is increased relative to the fuel rod region at the inner side while keeping the lattice pitch of the fuel assembly constant, thereby suppressing the increase of an absolute value of a void reactivity coefficient which tends to be caused when using MOX fuels as a fuel material, by utilizing neutron moderation due to a large quantity of coolants at the outer side of the fuel region. The void reactivity coefficient can be made substantially equal with that of uranium fuel assembly without greatly reducing a plutonium loading amount or without greatly increasing linear power density. (N.H.)

  3. Neutron disadvantage factors in heavy water and light water reactors

    International Nuclear Information System (INIS)

    Pop-Jordanov, J.

    1966-01-01

    A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

  4. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  5. 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)

  6. Experience with antimony activity removal process in Indian PHWRs

    International Nuclear Information System (INIS)

    Velmurugan, S.; Mittal, Vinit K.; Kumbhar, A.G.; Narasimhan, S.V.; Bhat, H.R.; Krishna Rao, K.S.; Upadhyay, S.K.; Jain, A.K.

    2008-01-01

    The problem of antimony (Sb) activity during decontamination was first encountered in NAPS-1 and Sb activity deposition took place during the decontamination resulting in poor decontamination factors (DF). Sb problem has been observed in PWRs and PHWRs elsewhere also. These utilities use an oxidative process involving the addition of H 2 O 2 to remove these Sb activities from the core and remove it on ion exchange resins. Experience in CANDU PHWRs indicated disappearance of H 2 O 2 in quantities higher than that observed in PWRs. This is attributed to the higher pick-up of H 2 O 2 by the magnetite/ferrites over large carbon steel surface present in the primary coolant system of PHWRs. Systematic work was carried out to understand the deposition of Sb on PHT system surfaces and a new method was evolved to remove the Sb activities from the system. This alternative reductive chemical process involve the addition of Nitrilo Tri Acetic Acid, Citric Acid and Rodine-92B and circulating the chemicals for a short period and then the Sb and other activities released from the core are removed by the mixed bed. Subsequent to the Sb removal process, the normal chemical decontamination of the system is carried out to remove 60 Co and other activities. This non-oxidizing Sb removal process was applied to NAPS-2 primary system prior to EMCCR. During this Sb removal process of NAPS-2, around 450 μCi/L activity of 124 Sb was released from the system surfaces to the formulation. Activity measurement in the samples collected and the on-line radiation field data indicated that deposition of Sb activities on system surfaces has been prevented by Rodine-92B and subsequently these activities have been removed by mixed bed IX columns. Antimony removal process worked successfully, but in the second normal decontamination process around 150 μCi/L activities came in the formulation which was not anticipated. As a result DF observed immediately after the decontamination campaign was not good

  7. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

    2010-06-01

    The value and importance of organizations in the nuclear industry engaged in the collection and analysis of operating experience and best practices has been clearly identified in various IAEA publications and exercises. Both facility safety and operational efficiency can benefit from such information sharing. Such sharing also benefits organizations engaged in the development of new nuclear power plants, as it provides information to assist in optimizing designs to deliver improved safety and power generation performance. In cooperation with Atomic Energy of Canada, Ltd, the IAEA organized the workshop on best practices in Heavy Water Reactor Operation in Toronto, Canada from 16 to 19 September 2008, to assist interested Member States in sharing best practices and to provide a forum for the exchange of information among participating nuclear professionals. This workshop was organized under Technical Cooperation Project INT/4/141, on Status and Prospects of Development for and Applications of Innovative Reactor Concepts for Developing Countries. The workshop participants were experts actively engaged in various aspects of heavy water reactor operation. Participants presented information on activities and practices deemed by them to be best practices in a particular area for consideration by the workshop participants. Presentations by the participants covered a broad range of operational practices, including regulatory aspects, the reduction of occupational dose, performance improvements, and reducing operating and maintenance costs. This publication summarizes the material presented at the workshop, and includes session summaries prepared by the chair of each session and papers submitted by the presenters

  8. Light-water reactor research and development

    International Nuclear Information System (INIS)

    1985-05-01

    This report on the national program of research and development on light water reactors is the second of two reports requested in 1982 by W. Kenneth Davis, Deputy Secretary of the Department of Energy. A first report, published in September 1983, treated the needs for safety-related R and D. In this second report, the Energy Research Advisory Board finds that, although many light water reactors are providing reliable and economic electricity, it appears unlikely that U.S. utilities will order additional reactors until the currently unacceptable economic risk, created by the regulatory climate and uncertain demand, is reduced. Thus it is unlikely that the private sector alone will fund major LWR design improvements. However, nuclear power will continue on its current course of expansion overseas. DOE participation is vitally needed to support the national interest in LWR technology. The report outlines R and D needs for a program to improve the safety, reliability, and economics of the present generation of plants; to develop evolutionary improved designs to be ready when needed; and to explore innovative longer-term concepts for deployment after the year 2000. The respective roles of government and the private sector are discussed

  9. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  10. Controlling hydrogen behavior in light water reactors

    International Nuclear Information System (INIS)

    Cullingford, H.S.; Edeskuty, F.J.

    1981-01-01

    In the aftermath of the incident at Three Mile Island Unit 2 (TMI-2), a new and different treatment of the Light Water Reactor (LWR) risks is needed for public safety because of the specific events involving hydrogen generation, transport, and behavior following the core damage. Hydrogen behavior in closed environments such as the TMI-2 containment building is a complex phenomenon that is not fully understood. Hence, an engineering approach is presented for prevention of loss of life, equipment, and environment in case of a large hydrogen generation in an LWR. A six-level defense strategy is described that minimizes the possibility of ignition of released hydrogen gas and otherwise mitigates the consequences of hydrogen release. Guidance is given to reactor manufacturers, utility companies, regulatory agencies, and research organizations committed to reducing risk factors and insuring safety of life, equipment, and environment

  11. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  12. 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

  13. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  14. Control of water chemistry in operating reactors

    International Nuclear Information System (INIS)

    Riess, R.

    1997-01-01

    Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ''modified'' B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs

  15. Control of water chemistry in operating reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riess, R [Siemens AG Unternehmensbereich KWU, Erlangen (Germany)

    1997-02-01

    Water chemistry plays a major role in fuel cladding corrosion and hydriding. Although a full understanding of all mechanisms involved in cladding corrosion does not exist, controlling the water chemistry has achieved quite some progress in recent years. As an example, in PWRs the activity transport is controlled by operating the coolant under higher pH-values (i.e. the ``modified`` B/Li-Chemistry). On the other hand, the lithium concentration is limited to a maximum value of 2 ppm in order to avoid an acceleration of the fuel cladding corrosion. In BWR plants, for example, the industry has learned on how to limit the copper concentration in the feedwater in order to limit CILC (Copper Induced Localized Corrosion) on the fuel cladding. However, economic pressures are leading to more rigorous operating conditions in power reactors. Fuel burnups are to be increased, higher efficiencies are to be achieved, by running at higher temperatures, plant lifetimes are to be extended. In summary, this paper will describe the state of the art in controlling water chemistry in operating reactors and it will give an outlook on potential problems that will arise when going to more severe operating conditions. (author). 3 figs, 6 tabs.

  16. Advanced ceramic cladding for water reactor fuel

    International Nuclear Information System (INIS)

    Feinroth, H.

    2000-01-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of approximately 60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies ge50% would be examined

  17. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  18. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  19. Some areas of concern in Indian PHWRs from regulatory perspective

    International Nuclear Information System (INIS)

    Gupta, V.K.

    1991-01-01

    The basic concern from regulatory perspective in the operation of Indian PHWRs is for radiation exposure to the occupational workers and to the members of public during normal operation as well as abnormal conditions. The radiation exposure to the occupational workers is the result of radiation conditions in the plant and the practices followed for operation and maintenance. Both technical and administrative actions are responsible in controlling the radiation exposures. As far as exposure to the members of public is concerned, integrity of heat transport and moderator systems, performance of the ventilation system and integrity of fuel cladding are important elements during normal operation and some of the anticipated operational occurrences. Containment systems play an important role in controlling the impact in public domain during accident conditions. Elaborate emergency preparedness plans ready in advance perfected and optimised through drills and exercises give an assurance that should a mishap occur requiring emergency action in the public domain, adequate and necessary actions to reduce the radiological consequences will be taken. In this context, four areas of interest are: Radiation Exposure of Occupational Workers, Fuel Performance, Containment Systems and Emergency Preparedness in Public Domain. (author)

  20. Some areas of concern in Indian PHWRs from regulatory perspective

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, V K [Operating Plants Safety Division, Atomic Energy Regulatory Board, Bhabha Atomic Research Centre (BARC), Bombay (India)

    1991-04-01

    The basic concern from regulatory perspective in the operation of Indian PHWRs is for radiation exposure to the occupational workers and to the members of public during normal operation as well as abnormal conditions. The radiation exposure to the occupational workers is the result of radiation conditions in the plant and the practices followed for operation and maintenance. Both technical and administrative actions are responsible in controlling the radiation exposures. As far as exposure to the members of public is concerned, integrity of heat transport and moderator systems, performance of the ventilation system and integrity of fuel cladding are important elements during normal operation and some of the anticipated operational occurrences. Containment systems play an important role in controlling the impact in public domain during accident conditions. Elaborate emergency preparedness plans ready in advance perfected and optimised through drills and exercises give an assurance that should a mishap occur requiring emergency action in the public domain, adequate and necessary actions to reduce the radiological consequences will be taken. In this context, four areas of interest are: Radiation Exposure of Occupational Workers, Fuel Performance, Containment Systems and Emergency Preparedness in Public Domain. (author)

  1. Is light water reactor technology sustainable?

    International Nuclear Information System (INIS)

    Rothwell, G.; Van der Zwaan, B.

    2001-01-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  2. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  3. Westinghouse Water Reactor Divisions quality assurance plan

    International Nuclear Information System (INIS)

    1977-09-01

    The Quality Assurance Program used by Westinghouse Water Reactor Divisions is described. The purpose of the program is to assure that the design, materials, and workmanship on Nuclear Steam Supply System (NSSS) equipment meet applicable safety requirements, fulfill the requirements of the contracts with the applicants, and satisfy the applicable codes, standards, and regulatory requirements. This program satisfies the NRC Quality Assurance Criteria, 10CFR50 Appendix B, to the extent that these criteria apply to safety related NSSS equipment. Also, it follows the regulatory position provided in NRC regulatory guides and the requirements of ANSI Standard N45.2.12 as identified in this Topical Report

  4. Instrument lance for boiling water reactors

    International Nuclear Information System (INIS)

    Proell, N.; Bertz, S.; Graebener, K.H.

    1980-01-01

    The instrument lance contains in the lance cover pipe a thimble as part of the drive chamber system. Other thimbles serve to carry neutron detectors. Detectors can be exchanged without opening the reactor pressure vessel and without removing the fuel elements. Furthermore the detector exchange is independent from the fuel element cycle. The measurement lance passes from the bottom of the pressure vessel over the total hight of the core in the water ducts between the fuel elements and can thus determine the neutron flux distribution. (DG) [de

  5. Light-water-reactor hydrogen manual

    International Nuclear Information System (INIS)

    Camp, A.L.; Cummings, J.C.; Sherman, M.P.; Kupiec, C.F.; Healy, R.J.; Caplan, J.S.; Sandhop, J.R.; Saunders, J.H.

    1983-06-01

    A manual concerning the behavior of hydrogen in light water reactors has been prepared. Both normal operations and accident situations are addressed. Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation. Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate. A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful. Different sections are written at different levels, according to the most likely audience. The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procedures

  6. Development trends in light water reactors

    International Nuclear Information System (INIS)

    Fogelstroem, L.; Simon, M.

    1988-01-01

    The present market for new nuclear power plants is weak, but is expected to pick up again, which is why great efforts are being made to further develop the light water reactor line for future applications. There is both a potential and a need for further improvement, for instance with respect to even higher cost efficiency, a simplified operating permit procedure, shorter construction periods, and increased operational flexibility to meet rising demands in load following behavior and in better cycle data of fuel elements. However, also public acceptance must not be forgotten when deciding about the line to be followed in the development of LWR technology. (orig.) [de

  7. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  8. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    Yoshio Matsuo

    1987-01-01

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  9. Is light water reactor technology sustainable?

    Energy Technology Data Exchange (ETDEWEB)

    Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

    2001-07-01

    This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

  10. Thermodynamic analysis of a supercritical water reactor

    International Nuclear Information System (INIS)

    Edwards, M.

    2007-01-01

    A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)

  11. Screening reactor steam/water piping systems for water hammer

    International Nuclear Information System (INIS)

    Griffith, P.

    1997-09-01

    A steam/water system possessing a certain combination of thermal, hydraulic and operational states, can, in certain geometries, lead to a steam bubble collapse induced water hammer. These states, operations, and geometries are identified. A procedure that can be used for identifying whether an unbuilt reactor system is prone to water hammer is proposed. For the most common water hammer, steam bubble collapse induced water hammer, six conditions must be met in order for one to occur. These are: (1) the pipe must be almost horizontal; (2) the subcooling must be greater than 20 C; (3) the L/D must be greater than 24; (4) the velocity must be low enough so that the pipe does not run full, i.e., the Froude number must be less than one; (5) there should be void nearby; (6) the pressure must be high enough so that significant damage occurs, that is the pressure should be above 10 atmospheres. Recommendations on how to avoid this kind of water hammer in both the design and the operation of the reactor system are made

  12. Generic safety issues for nuclear power plants with pressurized heavy water reactors and measures for their resolution

    International Nuclear Information System (INIS)

    2007-06-01

    be used in reassessing the safety of individual operating plants. In 1998, the IAEA completed IAEA-TECDOC-1044 entitled Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for their Resolution and established the associated LWRGSIDB database (Computer Manual Series No. 13). The present compilation, which is based on broad international experience, is an extension of this work to cover pressurized heavy water reactors (PHWRs). As in the case of LWRs, it is one element in the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It addresses generic safety issues identified in nuclear power plants using PHWRs. In most cases, the measures taken or planned to resolve these issues are also identified. The work on this report was initiated by the Senior Regulators of Countries Operating CANDU-Type Nuclear Power Plants at one of their annual meetings. It was carried out within the framework of the IAEA's programme on National Regulatory Infrastructure for Nuclear Installation Safety and serves to enhance regulatory effectiveness through the exchange of safety related information

  13. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Gruber, E.E.

    1996-07-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

  14. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  15. Carbon-14 in reactor plant water

    International Nuclear Information System (INIS)

    Knowles, G.K.

    1979-01-01

    The method for the analysis of 14 C in reactor plant water and various waste streams previously used at the Idaho National Engineering Laboratory has been shown to be ineffective for samples which contain organic compounds. The previous method consisted of acidification and refluxing of the sample, precipitation of the liberated CO 2 , and subsequent analysis by the liquid scintillation method. The method was simple but it did not convert all compounds containing 14 C in the sample to CO 2 . The new method, while it is based on the previous method, has been improved by employing a strong oxidant, potassium persulfate and silver nitrate, for more complete oxidation of the organics to CO 2 . The new method yields 14 C values that have typically been one to two orders of magnitude higher than the values obtained using the former method. This indicates that most of the 14 C present in the current reactor water samples being analyzed is associated with trace amounts of organics

  16. Water inventory management in condenser pool of boiling water reactor

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  18. NEPTUNE: a modular system for light-water reactor calculation

    International Nuclear Information System (INIS)

    Bouchard, J.; Kanevoky, A.; Reuss, P.

    1975-01-01

    A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

  19. Negative power coefficient on PHWRs with CARA fuel

    International Nuclear Information System (INIS)

    Lestani, H.A.; González, H.J.; Florido, P.C.

    2014-01-01

    Highlights: • A PHWR fuel was optimized to obtain a negative power coefficient. • Fuel cost, being a measure of design investment efficiency, was optimized. • Influence on power coefficient of geometrical and economical parameters’ was studied. • Different neutronic absorbers were studied; pure absorbers can be used. • Thermal and economical models were developed to complement neutronic assessment. - Abstract: A study of power coefficient of reactivity in heavy water reactors is made analyzing the reactivity components of fuels with several modifications oriented at reducing the coefficient. A cell model is used for neutronics calculations; a non-linear two dimensional model is used to evaluate the thermal changes that follow a power change; and a levelized unit energy cost model is used to assess the economical feasibility of the design changes introduced to reduce power coefficient. The necessity of modelling all the aforementioned quantities in a coupled scheme is stressed, as a strong interdependence was found. A series of design changes complied with a negative power coefficient of reactivity, with a feasible power radial distribution and with low refuelling cost. Some investigation lines that exceed the fuel cell study and deal with the plant operation are marked as potentially addressing the stable operation of big heavy water reactors

  20. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

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

  1. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  2. Procedure for operating a heavy water cooled power reactor

    International Nuclear Information System (INIS)

    Rau, P.; Kumpf, H.

    1981-01-01

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

  3. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  4. A review of fast reactor programme in India - April 1992

    Energy Technology Data Exchange (ETDEWEB)

    Paranjpe, S R [Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu (India)

    1992-07-01

    There is no change in the basic policy for development of nuclear energy in India. Fast Breeder Reactors are required to be available commercially to supply increasing quantities of nuclear energy when the first phase programme of deployment of Pressurised Heavy Water Reactors would be reaching the limit imposed by indigenously available natural uranium. Based on presently proven reserves of economically exploitable uranium one cannot expect to support more than 10 to 15 million kilowatt of installed capacity of PHWRs. The immediate goal of the Fast Reactor Programme therefore, remains completion by 2002-2003 of the first 500 MWe Prototype Fast Breeder Reactor which will become the first reactor in the series of reactors to be built there afterwards. This will enable addition of one 500 MWe reactor each year even if the first phase of programme of PHWR is limited to 6.0 million kilowatt. The capital cost of installed kilowatt for FBRs is expected to be comparable to the capital cost per kilowatt for PHWRS. It is expected to launch the construction of PFBR in the next 2 or 3 years as soon as the over all economic condition shows some improvement. In the meantime, manufacturing development of important NSS components like Steam Generators, Sodium Pumps, Main Vessel and Inner Vessel has been initiated. Detailed designs of Control Rod Drive Mechanism (Primary) has been completed and contacts with the manufacturers are being established to identify the industry which would be entrusted with the responsibility of manufacturing the Control Rod Drive Mechanisms. Manufacturing technology for making cladding tubes of D9 stainless steel has been developed and significant progress has been made towards the production of hexagonal wrapper (i.e. Hex-Cans). Inclined Fuel Transfer Machine for loading and unloading the fuel from the Main Vessel has been designed and manufacturing of the prototype machine has been initiated. It is hoped that these steps will enable timely completion

  5. Development of supercritical water reactors in Russia and abroad

    International Nuclear Information System (INIS)

    Glebov, A.P.; Klushin, A.V.

    2014-01-01

    The results of Russian and foreign studies on the water-cooled high critical parameters reactors are analyzed. Developments on this subject are conducted in more than 15 countries. The advantages of WWER- SCP and characteristics of experimental reactor of WWER-SCP-30 are discussed. It is noted that priority task is to develop a reactor with thermal neutron spectrum with a subsequent transition to the reactor with a fast neutron spectrum [ru

  6. Nuclear fuel for light water reactors

    International Nuclear Information System (INIS)

    Etemad, A.

    1976-01-01

    The goal of the present speech is to point out some of the now-a-day existing problems related to the fuel cycle of light water reactors and to foresee their present and future solutions. Economical aspects of nuclear power generation have been considerably improving, partly through technological advancements and partly due to the enlargement of unit capacity. The fuel cycle, defined in the course of this talk, discusses the exploration, mining, ore concentration, purification, conversion, enrichment, manufacturing of fuel elements, their utilization in a reactor, their discharge and subsequent storage, reprocessing, and their re-use or disposal. Uranium market in the world and the general policy of several uranium owning countries are described. The western world requirement for uranium until the year 2000, uranium resources and the nuclear power programs in the United States, Australia, Canada, South Africa, France, India, Spain, and Argentina are discussed. The participation of Iran in a large uranium enrichment plant based on French diffusion technology is mentioned

  7. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2005-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

  8. Hydrogen behavior in light-water reactors

    International Nuclear Information System (INIS)

    Berman, M.; Cummings, J.C.

    1984-01-01

    The Three Mile Island accident resulted in the generation of an estimated 150 to 600 kg of hydrogen, some of which burned inside the containment building, causing a transient pressure rise of roughly 200 kPa (2 atm). With this accident as the immediate impetus and the improved safety of reactors as the long-term goal, the nuclear industry and the Nuclear Regulatory Commission initiated research programs to study hydrogen behavior and control during accidents at nuclear plants. Several fundamental questions and issues arise when the hydrogen problem for light-water-reactor plants is examined. These relate to four aspects of the problem: hydrogen production; hydrogen transport, release, and mixing; hydrogen combustion; and prevention or mitigation of hydrogen combustion. Although much has been accomplished, some unknowns and uncertainties still remain, for example, the rate of hydrogen production during a degraded-core or molten-core accident, the rate of hydrogen mixing, the effect of geometrical structures and scale on combustion, flame speeds, combustion completeness, and mitigation-scheme effectiveness. This article discusses the nature and extent of the hydrogen problem, the progress that has been made, and the important unresolved questions

  9. Modified hot-conditioning of PHT system surfaces of PHWRs

    Energy Technology Data Exchange (ETDEWEB)

    Venkateswaran, G [Bhabha Atomic Research Centre, Trombay, Bombay (India)

    1997-02-01

    The increased awareness on the importance of controlling activity transport and radiation buildup on out-of-core surfaces of water cooled nuclear reactors is leading to a host of measures both from chemistry as well as engineering sides being undertaken. Passivation of the surfaces of structural materials is one such. Pressurised Heavy Water Reactors of CANDU design use large surface area of carbon steel alloy in the Primary Heat Transport System. Hot-conditioning of the PHT system with deoxygenated light water at temperatures {approx_equal} 473 - 523 K during commissioning stage is done to form a protective magnetite film on the surfaces of carbon steel essentially to guard this material from corrosion during the intervening period between initial commissioning and first fuel loading and achieving nuclear heat. However, a need is felt to improve the quality of this magnetite film and control the crud release so that the twin objectives of controlling the corrosion of carbon steel and reducing a possible deposition of corrosion products on surfaces of fuel clad could be achieved. Laboratory static autoclave investigations have been carried out on the formation of protective magnetite film on carbon steel at 473 K, pH 10 (pH at 298 K) deoxygenated aqueous solutions of chelants like HEDTA, DTPA, NTA apart from EDTA. Additionally, influence of AVT chemicals like hydrazine, cyclohexylamine, morpholine and additives like glucose, boric acid has been studied. The data have been compared with the standard procedure of hot-conditioning namely with simple LiOH. It is found that chelants increase the base metal loss but the oxide formed is more protective than the one formed under simple LiOH treatment. The efficiency of passivation is greatly enhanced by hydrazine and boric acid while it is adversely affected by glucose. AVT chemicals acts as effective corrosion inhibitors. (author). 14 refs, 2 figs, 4 tabs.

  10. Modified hot-conditioning of PHT system surfaces of PHWRs

    International Nuclear Information System (INIS)

    Venkateswaran, G.

    1997-01-01

    The increased awareness on the importance of controlling activity transport and radiation buildup on out-of-core surfaces of water cooled nuclear reactors is leading to a host of measures both from chemistry as well as engineering sides being undertaken. Passivation of the surfaces of structural materials is one such. Pressurised Heavy Water Reactors of CANDU design use large surface area of carbon steel alloy in the Primary Heat Transport System. Hot-conditioning of the PHT system with deoxygenated light water at temperatures ≅ 473 - 523 K during commissioning stage is done to form a protective magnetite film on the surfaces of carbon steel essentially to guard this material from corrosion during the intervening period between initial commissioning and first fuel loading and achieving nuclear heat. However, a need is felt to improve the quality of this magnetite film and control the crud release so that the twin objectives of controlling the corrosion of carbon steel and reducing a possible deposition of corrosion products on surfaces of fuel clad could be achieved. Laboratory static autoclave investigations have been carried out on the formation of protective magnetite film on carbon steel at 473 K, pH 10 (pH at 298 K) deoxygenated aqueous solutions of chelants like HEDTA, DTPA, NTA apart from EDTA. Additionally, influence of AVT chemicals like hydrazine, cyclohexylamine, morpholine and additives like glucose, boric acid has been studied. The data have been compared with the standard procedure of hot-conditioning namely with simple LiOH. It is found that chelants increase the base metal loss but the oxide formed is more protective than the one formed under simple LiOH treatment. The efficiency of passivation is greatly enhanced by hydrazine and boric acid while it is adversely affected by glucose. AVT chemicals acts as effective corrosion inhibitors. (author). 14 refs, 2 figs, 4 tabs

  11. Feedwater processing method in a boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Izumitani, M; Tanno, K

    1976-09-06

    The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

  12. The heavy water accountancy for research reactors in JAERI

    International Nuclear Information System (INIS)

    Yoshijima, Tetsuo; Tanaka, Sumitoshi; Nemoto, Denjirou

    1998-11-01

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

  13. Technological readiness of evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Juhn, P.E.

    1999-01-01

    Nuclear energy has evolved to a mature industry that supplies over 16% of the world's electricity, and it represents an important option for meeting the global energy demands of the coming century in an environmentally acceptable manner. New, evolutionary water cooled reactor designs that build on successful performance of predecessors have been developed; these designs have generally been guided by wishes to reduce cost, to improve availability and reliability, and to meet increasingly stringent safety objectives. These three aspects are important factors in what has been called technological readiness for an expanded deployment of nuclear power; a major increase in utilization of nuclear power will only occur if it is economically competitive, and meets safety expectations. To this end, the industry will also have to maintain or improve the public perception of nuclear power as a benign, economical and reliable energy source. (author)

  14. Radiation Protection at Light Water Reactors

    CERN Document Server

    Prince, Robert

    2012-01-01

    This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

  15. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  16. Light water reactor lower head failure analysis

    International Nuclear Information System (INIS)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L.

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response

  17. Reprocessing technology for present water reactor fuels

    International Nuclear Information System (INIS)

    McMurray, P.R.

    1977-01-01

    The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

  18. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  19. Corrosion problems in light water nuclear reactors

    International Nuclear Information System (INIS)

    Berry, W.E.

    1984-01-01

    The corrosion problems encountered during the author's career are reviewed. Attention is given to the development of Zircaloys and attendant factors that affect corrosion; the caustic and chloride stress corrosion cracking (SCC) of austenitic stainless steel steam generator tubing; the qualification of Inconel Alloy 600 for steam generator tubing and the subsequent corrosion problem of secondary side wastage, caustic SCC, pitting, intergranular attack, denting, and primary side SCC; and SCC in weld and furnace sensitized stainless steel piping and internals in boiling water reactor primary coolants. Also mentioned are corrosion of metallic uranium alloy fuels; corrosion of aluminum and niobium candidate fuel element claddings; crevice corrosion and seizing of stainless steel journal-sleeve combinations; SCC of precipitation hardened and martensitic stainless steels; low temperature SCC of welded austenitic stainless steels by chloride, fluoride, and sulfur oxy-anions; and corrosion problems experienced by condensers

  20. Light water reactor lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Chavez, S.A.; Thinnes, G.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)] [and others

    1993-10-01

    This document presents the results from a US Nuclear Regulatory Commission-sponsored research program to investigate the mode and timing of vessel lower head failure. Major objectives of the analysis were to identify plausible failure mechanisms and to develop a method for determining which failure mode would occur first in different light water reactor designs and accident conditions. Failure mechanisms, such as tube ejection, tube rupture, global vessel failure, and localized vessel creep rupture, were studied. Newly developed models and existing models were applied to predict which failure mechanism would occur first in various severe accident scenarios. So that a broader range of conditions could be considered simultaneously, calculations relied heavily on models with closed-form or simplified numerical solution techniques. Finite element techniques-were employed for analytical model verification and examining more detailed phenomena. High-temperature creep and tensile data were obtained for predicting vessel and penetration structural response.

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

    International Nuclear Information System (INIS)

    Symes, R.J.

    1983-01-01

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

  2. Fuel behavior in advanced water reactors

    International Nuclear Information System (INIS)

    Bolme, A.B.

    1996-01-01

    Fuel rod behavior of advanced pressurized water reactors under steady state conditions has been investigated in this study. System-80+ and Westinghouse Vantage-5 fuels have been considered as advanced pressurized water reactor fuels to be analyzed. The purpose of this study is to analyze the sensitivity of ditferent models and the effect of selected design parameters on the overall fuel behavior. FRAPCON-II computer code has been used for the analyses. Different modelling options of FRAPCON-II have also been considered in these analyses. Analyses have been performed in two main parts. In the first part, effects of operating conditions on fuel behavior have been investigated. First, fuel rod response under normal operating conditions has been analyzed. Then, fuel rod response to different fuel ratings has been calculated. In the second part, in order to estimate the effect of design parameters on fuel behavior, parametric analyses have been performed. In this part, the effects of initial gap thickness, as fabricated fuel density, and initial fill gas pressure on fuel behavior have been analyzed. The computations showed that both of the fuel rods used in this study operate within the safety limits. However, FRAPCON-II modelling options have been resulted in different behavior due to their modelling characteristics. Hence, with the absence of experimental data, it is difficult to make assesment for the best fuel parameters. It is also difficult to estimate error associated with the results. To improve the performance of the code, it is necessary to develop better experimental correlations for material properties in order to analyze the eftect ot considerably different design parameters rather than nominal rod parameters

  3. Heavy water moderated reactors advances and challenges

    International Nuclear Information System (INIS)

    Meneley, D.A.; Olmstead, R.A.; Yu, A.M.; Dastur, A.R.; Yu, S.K.W.

    1994-01-01

    Nuclear energy is now considered a key contributor to world electricity production, with total installed capacity nearly equal to that of hydraulic power. Nevertheless, many important challenges lie ahead. Paramount among these is gaining public acceptance: this paper makes the basic assumption that public acceptance will improve if, and only if, nuclear power plants are operated safely and economically over an extended period of time. The first task, therefore, is to ensure that these prerequisites to public acceptance are met. Other issues relate to the many aspects of economics associated with nuclear power, include capital cost, operation cost, plant performance and the risk to the owner's investment. Financing is a further challenge to the expansion of nuclear power. While the ability to finance a project is strongly dependent on meeting public acceptance and economic challenges, substantial localisation of design and manufacture is often essential to acceptance by the purchaser. The neutron efficient heavy water moderated CANDU with its unique tube reactor is considered to be particularly well qualified to respond to these market challenges. Enhanced safety can be achieved through simplification of safety systems, design of the moderator and shield water systems to mitigate severe accident events, and the increased use of passive systems. Economics are improved through reduction in both capital and operating costs, achieved through the application of state-of-the-art technologies and economy of scale. Modular features of the design enhance the potential for local manufacture. Advanced fuel cycles offer reduction in both capital costs and fuelling costs. These cycles, including slightly enriched uranium and low grade fuels from reprocessing plants can serve to increase reactor output, reduce fuelling cost and reduce waste production, while extending resource utilisation. 1 ref., 1 tab

  4. Computer simulation of the NASA water vapor electrolysis reactor

    Science.gov (United States)

    Bloom, A. M.

    1974-01-01

    The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.

  5. On the slimeless water operation in the RBMK type reactors

    International Nuclear Information System (INIS)

    Margulova, T.Kh.; Mamet, V.A.; Nikitina, I.S.; Karakhanyan, L.N.

    1988-01-01

    Water chemistry conditions of the operating RBMK-1000 and RBMK-1500 units are analysed. Inevitability of iron oxide deposits in RBMK-1000 and particularly in RBMK-1500 reactors is demonstrated. Organization of a new slimeless correcting water chemistry for RBMK-1000 and RBMK-1500 reactors is recommended

  6. Development of light water reactors and subjects for hereafter

    International Nuclear Information System (INIS)

    Murao, Yoshio

    1995-01-01

    As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

  7. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    Queiser, H.

    1976-01-01

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.) [de

  8. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  9. Suppression device for the reactor water level lowering

    International Nuclear Information System (INIS)

    Kasuga, Hajime; Kasuga, Hiroshi.

    1984-01-01

    Purpose: To suppress the lowering in the reactor water level so as to avoid unnecessary actuation of ECCS upon generation of transient changes which forecasts the lowering of the reactor water level in a BWR type reactor. Constitution: There are provided a water level suppression signal generator for generating a water level suppression signal upon generation of a transient change signal which forecasts the water level lowering in a nuclear reactor and a recycling flow rate controller that applies a recycling flow rate control signal to a recycling pump drive motor by the water level lowering suppression signal. The velocity of the recycling pump is controlled by a reactor scram signal by way of the water level lowering suppresion signal generator and a recycling flow rate controller. Then, the recycling reactor core flow rate is decreased and the void amount in the reactor is transiently increased where the water level tends to increase. Accordingly, the water level lowering by the scram is moderated by the increasing tendency of the water level. (Ikeda, J.)

  10. The steam generating heavy water reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

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

  11. Graphite-water steam-generating reactor in the USSR

    Energy Technology Data Exchange (ETDEWEB)

    Dollezhal, N A [AN SSSR, Moscow

    1981-10-01

    One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

  12. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  13. 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)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Uranium utilization of light water cooled reactors and fast breeders

    International Nuclear Information System (INIS)

    Stojadinovic, Timm

    1991-08-01

    The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

  16. Auxiliary water supply device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In the device of the present invention, a cooling condensation means is disposed to a steam discharge channel of a turbine for driving pumps to directly return condensates to the reactor, so that the temperature of the suppression pool water is not elevated. Namely, the cooling condensation means for discharged steams is disposed to the discharge channel of the turbine. The condensate channel from the cooling condensation means is connected to a sucking side of the turbine driving pump. With such a constitution, when the reactor is isolated from a main steam system, reactor scram is conducted. Although the reactor water level is lowered by the reactor scram, the lowering of the reactor water level is prevented by supplementing cooling water by the turbine driving pump using steams generated in the reactor as a power source. The discharged steams after driving the turbine are cooled and condensated by the cooling condensation means by way of the discharge channel and returned to the reactor again by way of the condensate channel. With such procedures, since the temperature of suppression pool water is not elevated, there is no need to operate other cooling systems. In addition, auxiliary water can be supplied for a long period of time. (I.S.)

  17. The effect of heavy water reactors and liquid fuel reactors on the long-term development of nuclear energy

    International Nuclear Information System (INIS)

    Brand, P.; Wiechers, W.K.

    1974-01-01

    The effects of the rates at which various combinations of power reactor types are installed on the long-range (to the year 2040) uranium and plutonium inventory requirements are examined. Consideration is given to light water reactors, fast breeder reactors, high temperature gas-cooled reactors, heavy water reactors, and thermal breeder reactors, in various combinations, and assuming alternatively a 3% and a 5% growth in energy demand

  18. Transmutation of Americium in Light and Heavy Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

    2009-06-15

    There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

  19. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  20. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  1. Alternative water injection device to reactor equipment facility

    International Nuclear Information System (INIS)

    Yamashita, Masahiro.

    1995-01-01

    The device of the present invention injects water to the reactor and the reactor container continuously for a long period of time for preventing occurrence of a severe accident in a BWR type reactor and maintaining the integrity of the reactor container even if the accident should occur. Namely, diesel-driven pumps disposed near heat exchangers of a reactor after-heat removing system (RHR) are operated before the reactor is damaged by the after heat to cause reactor melting. A sucking valve disposed to a pump sucking pipeline connecting a secondary pipeline of the RHR heat exchanger and the diesel driving pump is opened. A discharge valve disposed to a pump discharge pipeline connecting a primary pipeline of the RHR heat exchanger and the diesel driving pump is opened. With such procedures, sea water is introduced from a sea water taking port through the top end of the secondary pipeline of the RHR heat exchanger and water is injected into the inside of the pressure vessel or the reactor container by way of the primary pipeline of the RHR heat exchanger. As a result, the reactor core is prevented from melting even upon occurrence of a severe accident. (I.S.)

  2. Calculation of photon dose for Dalat research reactor in case of loss of reactor tank water

    International Nuclear Information System (INIS)

    Le Vinh Vinh; Huynh Ton Nghiem; Nguyen Kien Cuong

    2007-01-01

    Photon sources of actinides and fission products were estimated by ORIGEN2 code with the modified cross-section library for Dalat research reactor (DRR) using new cross-section generated by WIMS-ANL code. Photon sources of reactor tank water calculated from the experimental data. MCNP4C2 with available non-analog Monte Carlo model and ANSI/ANL-6.1.1-1977 flux-to-dose factors were used for dose estimation. The agreement between calculation results and those of measurements showed that the methods and models used to get photon sources and dose were acceptable. In case the reactor water totally leaks out from the reactor tank, the calculated dose is very high at the top of reactor tank while still low in control room. In the reactor hall, the operation staffs can access for emergency works but with time limits. (author)

  3. Grey water treatment in UASB reactor at ambient temperature.

    Science.gov (United States)

    Elmitwalli, T A; Shalabi, M; Wendland, C; Otterpohl, R

    2007-01-01

    In this paper, the feasibility of grey water treatment in a UASB reactor was investigated. The batch recirculation experiments showed that a maximum total-COD removal of 79% can be obtained in grey-water treatment in the UASB reactor. The continuous operational results of a UASB reactor treating grey water at different hydraulic retention time (HRT) of 20, 12 and 8 hours at ambient temperature (14-24 degrees C) showed that 31-41% of total COD was removed. These results were significantly higher than that achieved by a septic tank (11-14%), the most common system for grey water pre-treatment, at HRT of 2-3 days. The relatively lower removal of total COD in the UASB reactor was mainly due to a higher amount of colloidal COD in the grey water, as compared to that reported in domestic wastewater. The grey water had a limited amount of nitrogen, which was mainly in particulate form (80-90%). The UASB reactor removed 24-36% and 10-24% of total nitrogen and total phosphorus, respectively, in the grey water, due to particulate nutrients removal by physical entrapment and sedimentation. The sludge characteristics of the UASB reactor showed that the system had stable performance and the recommended HRT for the reactor is 12 hours.

  4. Introduction to reactor internal materials for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

  5. Introduction to reactor internal materials for pressurized water reactor

    International Nuclear Information System (INIS)

    Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

    1994-06-01

    This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

  6. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  7. Auxiliary equipment for cooling water in a reactor

    International Nuclear Information System (INIS)

    Konno, Yasuhiro; Sakairi, Toshiaki.

    1975-01-01

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

  8. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  9. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-09-01

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

  10. Material and water chemistry for a ferritic reactor coolant system in pressure water reactors

    International Nuclear Information System (INIS)

    Stieding, L.

    1979-04-01

    The use of unplated, low-alloy steels in a boric acid controlled PWR is not considered possible without changing the water conditions during the start-up and shut-down periods of the reactor. The significant pH reduction of the water due to boric acid during these periods most probably leads to damage of the magnetite protective layers followed by selective corrosion. As this highly important process has not been sufficiently evaluated with respect to our specific application problem, more detailed information will be necessary. KWU test facilities provide a means of performing such tests. In order to avoid corrosion attack during the above operating conditions, an inhibition of the water with 7 Li-borate is recommended which, however, will amount to approx. DM 60.000,-- per period of use. (orig.) [de

  11. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Park, J.Y.; Ruther, W.E.; Kassner, T.F.; Shack, W.J.

    1990-12-01

    Topics that have been investigated during this year include (1) SCC of A533-Gr B steel used in steam generator and reactor pressure vessels, (2) fatigue of Type 316NG SS, and (3) SCC of Type 347 and CF-3 cast duplex stainless steels in simulated BWR water. Crack-growth-rate (CGR) tests were performed on a composite A533-Gr B/Inconel-182 specimen in which the stress corrosion crack in the Inconel-182 weld metal penetrated and grew into the A533-Gr B steel. CGR tests were also conducted on conventional (unplated) and nickel- or gold-plated A533-Gr B specimens to provide insight into whether the nature of the surface layer on the low-alloy steel, either oxide corrosion products or a noble metal, influences the overall SCC process. CGR data on the A533-Gr B specimens were compared with the fatigue crack reference curves in the ASME Boiler and Pressure Vessel Code, Section XI, Appendix A. Fatigue tests were conducted on Type 316NG SS in air and simulated BWR water at low strain ranges and frequencies to better establish margins in the ASME Code Section III Fatigue Design Curves. CGR tests were also conducted on specimens of Type 347 SS with different heat-treatment conditions, and a specimen of CF-3 cast stainless steel with a ferrite content of 15.6%. The results were compared with previous data on another heat of Type 347 SS, which was very resistant to SCC, and a CF-3M steel with a ferrite content of 5%. 37 refs., 15 figs., 8 tabs

  12. RETRAN sensitivity studies of light water reactor transients. Final report

    International Nuclear Information System (INIS)

    Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

    1977-06-01

    This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

  13. Reactivity requirements and safety systems for heavy water reactors

    International Nuclear Information System (INIS)

    Kati, S.L.; Rustagi, R.S.

    1977-01-01

    The natural uranium fuelled pressurised heavy water reactors are currently being installed in India. In the design of nuclear reactors, adequate attention has to be given to the safety systems. In recent years, several design modifications having bearing on safety, in the reactor processes, protective and containment systems have been made. These have resulted either from new trends in safety and reliability standards or as a result of feed-back from operating reactors of this type. The significant areas of modifications that have been introduced in the design of Indian PHWR's are: sophisticated theoretical modelling of reactor accidents, reactivity control, two independent fast acting systems, full double containment and improved post-accident depressurisation and building clean-up. This paper brings out the evolution of design of safety systems for heavy water reactors. A short review of safety systems which have been used in different heavy water reactors, of varying sizes, has been made. In particular, the safety systems selected for the latest 235 MWe twin reactor unit station in Narora, in Northern India, have been discussed in detail. Research and Development efforts made in this connection are discussed. The experience of design and operation of the systems in Rajasthan and Kalpakkam reactors has also been outlined

  14. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    Mattern, J.

    1976-01-01

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

  15. Possibility of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-01-01

    The review of metal uranium properties including irradiation in the reactor core lead to the following conclusions. Using metal uranium in the heavy water reactors would be favourable from economic point of view for ita high density, i.e. high conversion factor and low cost of fuel elements fabrication. Most important constraint is swelling during burnup and corrosion

  16. Calculations for accidents in water reactors during operation at power

    International Nuclear Information System (INIS)

    Blanc, H.; Dutraive, P.; Fabrega, S.; Millot, J.P.

    1976-07-01

    The behaviour of a water reactor on an accident occurring as the reactor is normally operated at power may be calculated through the computer code detailed in this article. Reactivity accidents, loss of coolant ones and power over-running ones are reviewed. (author)

  17. Overview of environmental materials degradation in light-water reactors

    International Nuclear Information System (INIS)

    Shaaban, H.I.; Wu, P.

    1986-08-01

    This report provides a brief overview of analyses and conclusions reported in published literature regarding environmentally induced degradation of materials in operating light-water reactors. It is intended to provide a synopsis of subjects of concern rather than to address a licensing basis for any newly discovered problems related to reactor materials

  18. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

    Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

  19. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  20. Challenges in design of zirconium alloy reactor components

    International Nuclear Information System (INIS)

    Kakodkar, Anil; Sinha, R.K.

    1992-01-01

    Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

  1. Brief summary of water reactor fuel activities in China

    Energy Technology Data Exchange (ETDEWEB)

    Zhongyue, Zhang [China Inst. of Atomic Energy (CIAE), Beijing (China)

    1997-12-01

    The presentation briefly reviews the water reactor fuel activities in China describing: nuclear power development program and growth forecast; fuel performance;fuel performance code improvement; research and development plans. 1 ref., 3 figs, 2 tabs.

  2. Experience in operation of heavy water reactors

    International Nuclear Information System (INIS)

    Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

    1999-01-01

    The paper presents the main topics of the CANDU owners group (COG) meeting held in Mangalia, Romania on 7-10 September 1998. These meetings are part of the IAEA program for exchange of information related mainly to CANDU reactor operation safety. The first meeting for PHWR reactors took place in Vienna in 1989, followed by those in Argentina (1991), India (1994) and Korea (1996). The topics discussed at the meeting in Romania were: operation experience and recent major events, performances of CANDU reactors and safe operation, nuclear safety and operation procedures of PHWR, programs and strategies of lifetime management of installations and components of NPPs, developments and updates

  3. Water and Regolith Shielding for Surface Reactor Missions

    Science.gov (United States)

    Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

  4. Water and Regolith Shielding for Surface Reactor Missions

    International Nuclear Information System (INIS)

    Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

  5. Safety considerations concerning light water reactors in Sweden

    International Nuclear Information System (INIS)

    Nilsson, T.

    1977-01-01

    In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

  6. 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

  7. Analysis of an accelerator-driven subcritical light water reactor

    International Nuclear Information System (INIS)

    Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

    1997-01-01

    An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

  8. Computer modelling of eddy current probes for ISI of pressure tube/calandria tube assemblies in PHWRs

    International Nuclear Information System (INIS)

    Rao, B.P.C.; Shyamsunder, M.T.; Bhattacharya, D.K.; Raj, Baldev

    1992-01-01

    Non-destructive Evaluation (NDE) plays a major role in ensuring the safe and reliable operation of PHWRs which are the mainstay of India's nuclear power programme. An important in-service inspection (ISI) requirement in these reactors is carried out through Eddy Current Testing (ECT) of the pressure tube (PT)/calandria tube (CT) assemblies. The material of construction of these assemblies is zircaloy-2. The two main objectives of this ISI are the detection of garter spring between CT and PT and the profiling of gap between CT and PT. The paper discusses the work carried out at the authors' laboratory on the development of ECT probes for ISI of PT/CT assemblies. Emphasis has been given on the work done on the design and optimisation of the probes using computer modeling. A 2-D finite element code has been developed for this purpose. The code is developed around a diffusion equation which can be derived from Maxwell's equations governing the electromagnetic phenomenon. An axisymmetry has been considered, since the probes are bobbin type. Results of impedance plane outputs obtained by modelling and those by experiments using actual probes have shown good matching. Salient features of an indigenously developed interactive PC based data acquisition, analysis and retrieval system to cater to ISI of PC/CT assemblies are described. (author). 10 refs., 7 figs

  9. The Steam Generating Heavy Water Reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

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

  10. Pressurised water reactor fuel management using PANTHER

    International Nuclear Information System (INIS)

    Parks, G.T.; Knight, M.P.

    1996-01-01

    This paper describes the integration of Nuclear Electric's reactor physics code PANTHER with an automatic optimisation procedure designed to search for optimal PWR reload cores and assesses its performance. (Author)

  11. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  12. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  13. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

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

  14. Accumulated dose calculations in Indian PHWRs under DBA

    International Nuclear Information System (INIS)

    Nesaraj, David; Pradhan, A.S.; Bhardwaj, S.A.

    1996-01-01

    Accumulated gamma dose inside reactor building due to release of fission products from equilibrium core of Indian PHWR under accident condition has been assessed. The assessment has been done for the radiation tolerance limit of the critical equipment inside reactor building. The basic source data has been generated using computer code ORIGEN2 written and developed by Oak Ridge National Laboratory, USA (ORNL). This paper discusses the details of the calculations done on the basis of certain assumption which are mentioned at relevant places. The results indicate accumulated gamma dose at a few typical locations inside reactor building under accident condition. (author). 1 ref., 1 tab., 1 fig

  15. Device for controlling water supply to nuclear reactor

    International Nuclear Information System (INIS)

    Iwasaki, Toshio.

    1974-01-01

    Object: To smoothly control automatic water supply for realizing stable operation of a nuclear reactor by providing a flow rate limiting signal selection circuit and a preferential circuit in a water supply control device for a nuclear reactor wherein the speed of a recirculation pump may be changed in two-steps. Structure: Opening angle signals for a water supply regulating valve are controlled by a nuclear reactor water level signal, a vapor flow rate signal and a supplied water flow rate signal through an adder and an adjuster in response to a predetermined water level setting signal. When the water in the reactor is maintained at a predetermined level, a selection circuit receives a water pump condition signal for selecting one of the signals from a supplied water rate limiting signal generator generating signals for indicating whether one or two water supply pumps are operated. A low value preferential circuit passes the lower of the values generated from the selection circuit and the adder. The selection circuit receives a recirculation pump condition signal and selects either one of the signals from the supplied water flow rate limiting signal generator operated at high speed or low speed. A high value preferential circuit passes the higher value

  16. Heavy water leak detection using diffusion sampler

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-12-01

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

  18. Effect of water impurities on stress corrosion cracking in a boiling water reactor

    International Nuclear Information System (INIS)

    Ljungbery, L.G.; Cubicciotti, D

    1985-01-01

    A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

  19. Device for preventing cooling water from flowing out of reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  20. Analysis of thermal fatigue events in light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

    2000-09-01

    Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

  1. Sea water take-up facility for cooling reactor auxiliary

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. Automatic radiometric analyzer for nuclides in nuclear reactor water

    International Nuclear Information System (INIS)

    Kitamura, Masao; Tokoi, Hiromi; Kitaguchi, Hiroshi; Ozawa, Yoshihiro; Urata, Megumu.

    1981-01-01

    Purpose: To shorten the processing time and improve the accuracy for processing water sampled from reactor coolants, as well as simplify the mechanism of the apparatus. Constitution: Reactor water sampled from reactor coolants, after filtered out with insoluble solids, is stored in an ion exchange container. Thereafter, the amount of ion exchanged water is regulated by the coarse measurement of radioactivity concentration by a monitor. Further, ion exchange resins are charged from a resin tank, agitated by gases and dispersed into sampled water. Then, all of the radioactive iodines contained in the sample are collected in the resins. The resins are recovered through evacuation into instrumenting vessels for measurement of radioactivity. Since ion exchange resins are dispersed in the sampled water in this system, the processing time can be shortened. (Ikeda, J.)

  3. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    PetrusTakaki, N.

    2012-01-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  4. Light Water Reactor Sustainability Accomplishments Report

    Energy Technology Data Exchange (ETDEWEB)

    McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-02-01

    Welcome to the 2014 Light Water Reactor Sustainability (LWRS) Program Accomplishments Report, covering research and development highlights from 2014. The LWRS Program is a U.S. Department of Energy research and development program to inform and support the long-term operation of our nation’s commercial nuclear power plants. The research uses the unique facilities and capabilities at the Department of Energy national laboratories in collaboration with industry, academia, and international partners. Extending the operating lifetimes of current plants is essential to supporting our nation’s base load energy infrastructure, as well as reaching the Administration’s goal of reducing greenhouse gas emissions to 80% below 1990 levels by the year 2050. The purpose of the LWRS Program is to provide technical results for plant owners to make informed decisions on long-term operation and subsequent license renewal, reducing the uncertainty, and therefore the risk, associated with those decisions. In January 2013, 104 nuclear power plants operated in 31 states. However, since then, five plants have been shut down (several due to economic reasons), with additional shutdowns under consideration. The LWRS Program aims to minimize the number of plants that are shut down, with R&D that supports long-term operation both directly (via data that is needed for subsequent license renewal), as well indirectly (with models and technology that provide economic benefits). The LWRS Program continues to work closely with the Electric Power Research Institute (EPRI) to ensure that the body of information needed to support SLR decisions and actions is available in a timely manner. This report covers selected highlights from the three research pathways in the LWRS Program: Materials Aging and Degradation, Risk-Informed Safety Margin Characterization, and Advanced Instrumentation, Information, and Control Systems Technologies, as well as a look-ahead at planned activities for 2015. If you

  5. On the estimation of channel power distribution for PHWRs (Paper No. HMT-66-87)

    International Nuclear Information System (INIS)

    Parikh, M.V.; Kumar, A.N.; Krishnamohan, B.; Bhaskara Rao, P.

    1987-01-01

    In the case of PHWRs the estimation of channel power distribution is an important safety criteria. In this paper two methods based on theoretical estimation and the measured parameter are described. The comparison made shows good agreement in the prediction of channel power by both the methods. A parametric study in one of the measured parameters is also made which gives better agreement in results obtained. (author). 3 tabs

  6. Methodologies for assessment of the service life of pressure tubes in Indian PHWRs

    International Nuclear Information System (INIS)

    Sinha, R.K.; Sharma, A.; Madhusoodanan, K.; Sinha, S.K.; Malshe, U.D.

    1997-01-01

    For estimating safe service life of pressure tubes in Indian PHWRs, analytical methodologies have been developed to evaluate creep deformation, deuterium pick-up rate, blister growth at cold spot, and operating domain required for achieving leak-before-break. The paper provides an overview of these methodologies, and results of some studies carried out towards evolution of proposed fitness-for-service criteria for a pressure tube in contact with its calandria tube. (author)

  7. Development of advanced boiling water reactor for medium capacity

    International Nuclear Information System (INIS)

    Kazuo Hisajima; Yutaka Asanuma

    2005-01-01

    This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

  8. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

  9. Pressurized heavy-water reactor safety

    International Nuclear Information System (INIS)

    Pease, L.; Wilson, R.

    1977-09-01

    CANDU-PWR type reactors routinely release small amounts of radioactive liquids and gases and large quantities of low-grade waste heat. Radioactive emissions are usually below 1% of the derived release limits based on ICRP limits. Waste heat is common to all power plants and is not foreseen as a problem in Canadian conditions. Risk analysis shows a very low accident probability for CANDU type reactors. Multiple barriers to release of radionuclides, quality assurance, control, and inspection, containment systems, the shutdown system, the ECCS, and leak-before-break design, would all combine to mitigate the effects of an accident. (E.C.B.)

  10. Method of controlling power of a heavy water reactor

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1975-01-01

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

  11. Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (LWBR development program)

    International Nuclear Information System (INIS)

    Hecker, H.C.; Freeman, L.B.

    1981-08-01

    This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications

  12. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bahri, Che Nor Aniza Che Zainul, E-mail: anizazainul@gmail.com; Majid, Amran Ab.; Al-Areqi, Wadeeah M. [Nuclear Science Program, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, 43600 UKM Bangi, Selangor (Malaysia)

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  13. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  14. Flow analysis in a supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Oh, C.H.; Kochan, R.J.; Beller, J.M.

    1996-01-01

    Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

  15. Safety of light water reactors. Risks of nuclear technology

    International Nuclear Information System (INIS)

    Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

    2012-01-01

    The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

  16. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  17. Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-10-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

  18. Biological mine water treatment operating a one stage reactor system

    CSIR Research Space (South Africa)

    Baloyi, MJ

    2006-05-01

    Full Text Available rumen fluid as source of the fermentation organisms, were utilised as electron donor when sulphate, as the electron acceptor, is converted to sulphide. The feed water entered the reactor at the top, to allow the water to get in contact with grass...

  19. The experimental program of neutronphysics for advanced water reactors

    International Nuclear Information System (INIS)

    Martin-Deider, L.; Cathalu, S.; Santamarina, A.; Gomit, M.

    1985-11-01

    The C.E.A. and E.D.F. has jointly undertaken a program of experimental studies on under-moderated water lattices, with mixed oxide fuel UO 2 -PuO 2 . Undermoderated lattices offer high conversion ratios. This type of lattice could limit in the future the natural uranium consumption of pressurized water reactors. This experimental program is aimed at qualifying neutron transport calculations in a large range of moderating ratio (between 0.5 and 1.5). It includes three experiments: ERASME, a critical experiment of large size in the EOLE reactor at Cadarache; ICARE, an irradiation experiment in the MELUSINE reactor at Grenoble; and an experiment to measure the reactivity effects by oscillations in the MINERVE reactor at Cadarache [fr

  20. LOFA analyses for the water and helium cooled SEAFP reactors

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. Passive safety features in current and future water cooled reactors

    International Nuclear Information System (INIS)

    1990-11-01

    Better understanding of the passive safety systems and components in current and future water-cooled reactors may enhance the safety of present reactors, to the extend passive features are backfitted. This better understanding should also improve the safety of future reactors, which can incorporate more of these features. Passive safety systems and components may help to prevent accidents, core damage, or release radionuclides to the environment. The Technical Committee Meeting which was hosted by the USSR State Committee for Utilization of Nuclear Energy was attended by about 80 experts from 16 IAEA Member States and the NEA-OECD. A total of 21 papers were presented during the meeting. The objective of the meeting was to review and discuss passive safety systems and features of current and future water cooled reactor designs and to exchange information in this area of activity. A separate abstract was prepared for each of the 21 papers published in this proceedings. Refs, figs and tabs

  2. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    International Nuclear Information System (INIS)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-01-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean/US/laboratory/university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program

  3. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  4. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  5. Structural integrity of water reactor pressure boundary components

    International Nuclear Information System (INIS)

    Loss, F.J.

    1977-01-01

    The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

  6. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  7. 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

  8. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  9. Review of irradiation experiments for water reactor safety research

    International Nuclear Information System (INIS)

    Tobioka, Toshiaki

    1977-02-01

    A review is made of irradiation experiments for water reactor safety research under way in both commercial power plants and test reactors. Such experiments are grouped in two; first, LWR fuel performance under normal and abnormal operating conditions, and second, irradiation effects on fracture toughness in LWR vessels. In the former are fuel densification, swelling, and the influence of power ramp and cycling on fuel rod, and also fuel rod behavior under accident conditions in in-reactor experiment. In the latter are the effects of neutron exposure level on the ferritic steel of pressure vessels, etc.. (auth.)

  10. Nuclear powerplant standardization: light water reactors. Volume 2. Appendixes

    International Nuclear Information System (INIS)

    1981-06-01

    This volume contains working papers written for OTA to assist in preparation of the report, NUCLEAR POWERPLANT STANDARDIZATION: LIGHT WATER REACTORS. Included in the appendixes are the following: the current state of standardization, an application of the principles of the Naval Reactors Program to commercial reactors; the NRC and standardization, impacts of nuclear powerplant standardization on public health and safety, descriptions of current control room designs and Duke Power's letter, Admiral Rickover's testimony, a history of standardization in the NRC, and details on the impact of standardization on public health and safety

  11. Possibilities of using metal uranium fuel in heavy water reactors

    International Nuclear Information System (INIS)

    Djuric, B.; Mihajlovic, A.; Drobnjak, Dj.

    1965-11-01

    There are serious economic reasons for using metal uranium in heavy water reactors, because of its high density, i.e. high conversion factor, and low cost of fuel elements production. Most important disadvantages are swelling at high burnup and corrosion risk. Some design concepts and application of improved uranium obtained by alloying are promising for achievement of satisfactory stability of metal uranium under reactor operation conditions [sr

  12. Natural uranium fueled light water moderated breeding hybrid power reactors

    International Nuclear Information System (INIS)

    Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.; Levin, P.

    The feasibility of fission-fusion hybrid reactors based on breeding light water thermal fission systems is investigated. The emphasis is on fuel-self-sufficient (FSS) hybrid power reactors that are fueled with natural uranium. Other LWHRs considered include FSS-LWHRs that are fueled with spent fuel from LWRs, and LWHRs which are to supplement LWRs to provide a tandem LWR-LWHR power economy that is fuel-self-sufficient

  13. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  14. Controllability studies for an advanced CANDU boiling light water reactor

    International Nuclear Information System (INIS)

    Lepp, R.M.; Hinds, H.W.

    1976-12-01

    Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

  15. Method of operating a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Lysell, G.

    1975-01-01

    When operating a water-cooled nuclear reactor, in which the fuel rods consist of zirconium alloy tubes containing an oxidic nuclear fuel, stress corrosion in the tubes can be reduced or avoided if the power of the reactor is temporarily increased so much that the thermal expansion of the nuclear fuel produces a flow of the material in the tube. After that temporary power increase the power output is reduced to the normal power

  16. UK methods for studying fuel management in water moderated reactors

    International Nuclear Information System (INIS)

    Fayers, F.J.

    1970-10-01

    Current UK methods for studying fuel management and for predicting the reactor physics performance for both light and heavy water moderated power reactors are reviewed. Brief descriptions are given of the less costly computer codes used for initial assessment studies, and also the more elaborate programs associated with detailed evaluation are discussed. Some of the considerations influencing the accuracy of predictions are included with examples of various types of experimental confirmation. (author)

  17. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  18. Dual pressurized light water reactor producing 2000 M We

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

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

  19. A `big-mac` high converting water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ronen, Y; Dali, Y [Ben-Gurion Univ. of the Negev, Beersheba (Israel). Dept. of Nuclear Engineering

    1996-12-01

    Currently an effort is being made to get rid of plutonium. Therefore, at this time, a scientific study of a high converting reactor seems to be out of place. However , it is our opinion that the future of nuclear energy lies, among other things in the clever utilization of plutonium. It is also our opinion that one of the best ways to utilize plutonium is in high converting water reactors (authors).

  20. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  1. Entropy Generation Minimization for Reverse Water Gas Shift (RWGS Reactors

    Directory of Open Access Journals (Sweden)

    Lei Zhang

    2018-05-01

    Full Text Available Thermal design and optimization for reverse water gas shift (RWGS reactors is particularly important to fuel synthesis in naval or commercial scenarios. The RWGS reactor with irreversibilities of heat transfer, chemical reaction and viscous flow is studied based on finite time thermodynamics or entropy generation minimization theory in this paper. The total entropy generation rate (EGR in the RWGS reactor with different boundary conditions is minimized subject to specific feed compositions and chemical conversion using optimal control theory, and the optimal configurations obtained are compared with three reference reactors with linear, constant reservoir temperature and constant heat flux operations, which are commonly used in engineering. The results show that a drastic EGR reduction of up to 23% can be achieved by optimizing the reservoir temperature profile, the inlet temperature of feed gas and the reactor length simultaneously, compared to that of the reference reactor with the linear reservoir temperature. These optimization efforts are mainly achieved by reducing the irreversibility of heat transfer. Optimal paths have subsections of relatively constant thermal force, chemical force and local EGR. A conceptual optimal design of sandwich structure for the compact modular reactor is proposed, without elaborate control tools or excessive interstage equipment. The results can provide guidelines for designing industrial RWGS reactors in naval or commercial scenarios.

  2. Study of plutonium recycling physics in light water reactors

    International Nuclear Information System (INIS)

    Reuss, Paul

    1979-10-01

    A stock of plutonium from the reprocessing of thermal neutron reactor fuel is likely to appear in the next few years. The use of this plutonium as fuel replacing 235 U in thermal reactors is probably more interesting than simple stock-piling storage: immobilization of a capital which moreover would deteriorate by radioactive decay of isotope 241 also fissile and present to an appreciable extend in plutonium from reprocessing (half-life 15 years); recycling, on the other hand, will supply energy without complete degradation of the stock for fast neutron reactor loads, the burned matter having been partially renewed by conversion; furthermore the use of plutonium will meet the needs created by a temporary pressure on the naturel and/or enriched uranium market. For these two reasons the recycling of plutonium in thermal neutron reactors is being considered seriously today. The present work is confined to neutronic aspects and centres mainly on pressurized water-moderated reactors, the most highly developed at present in France. Four aspects of the problem are examined: 1. the physics of a plutonium-recycling reactor special features of neutronic phenomena with respect to the 'conventional' scheme of the 235 U burning reactor; 2. calculation of a plutonium-recycling reactor: adaptation of standard methods; 3. qualification of these calculations from the viewpoint of both data and inevitable approximations; 4. the fuel cycle and particularly the equivalence of fissile matters [fr

  3. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  4. A water inner circulation device for a reactor vessel

    International Nuclear Information System (INIS)

    Eriksson, O.

    1976-01-01

    A water inner circulation device for a reactor vessel comprising a pump mounted in the reactor vessel and driven by a water-cooled electric motor mounted in a housing outside the reactor vessel, the shaft of the pump passing through the reactor-vessel bottom and being coupled to the motor shaft in a member mechanically connected to the bottom of the reactor vessel in the vicinity of the motor housing, the pump shaft being surrounded by a resilient sealing ring, the reactor vessel communicating with the cooling channels of the pump, when the latter is operating, via a slot surrounding the pump hollow cylindrical shaft, characterized in that the slot inner end is used for/forming a circular space surrounding the pump shaft and surrounded by the motorhousing, in which is coaxially mounted a separating cylindral wall, the upper edge of which is tightly applied against the inner wall of the motor-housing to which it is fastened vertically, the inner surface of said wall being turned towards the outer surface of a circular packing-box, the outer surface of said separating wall constituting a separating radical inner surface for a circular chamber through which flow the motor cooling water. (author)

  5. Gamma spectroscopy in water cooled reactors

    International Nuclear Information System (INIS)

    Persault, M.

    1977-10-01

    Gamma spectroscopy analysis of spent fuels in power reactors; study of two typical cases: determination of the power distribution by the mean of the activity of a low periodic element (Lanthanum 140) and determination of the burnup absolute rate by examining the ratio of Cesium 134 and Cesium 137 activities. Measures were realized on fuel solutions and on fuel assemblies. Development of a power distribution map of the assemblies and comparison with the results of a three dimensional calculation of core evolution [fr

  6. Spiral-shaped reactor for water disinfection

    KAUST Repository

    Soukane, Sofiane; Ait-Djoudi, Fariza; Naceur, Wahib M.; Ghaffour, NorEddine

    2016-01-01

    Chlorine-based processes are still widely used for water disinfection. The disinfection process for municipal water consumption is usually carried out in large tanks, specifically designed to verify several hydraulic and disinfection criteria

  7. 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

  8. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  9. Novel Photocatalytic Reactor Development for Removal of Hydrocarbons from Water

    Directory of Open Access Journals (Sweden)

    Morgan Adams

    2008-01-01

    Full Text Available Hydrocarbons contamination of the marine environment generated by the offshore oil and gas industry is generated from a number of sources including oil contaminated drill cuttings and produced waters. The removal of hydrocarbons from both these sources is one of the most significant challenges facing this sector as it moves towards zero emissions. The application of a number of techniques which have been used to successfully destroy hydrocarbons in produced water and waste water effluents has previously been reported. This paper reports the application of semiconductor photocatalysis as a final polishing step for the removal of hydrocarbons from two waste effluent sources. Two reactor concepts were considered: a simple flat plate immobilised film unit, and a new rotating drum photocatalytic reactor. Both units proved to be effective in removing residual hydrocarbons from the effluent with the drum reactor reducing the hydrocarbon content by 90% under 10 minutes.

  10. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  11. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-01-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  12. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  13. The higher order flux mapping method in large size PHWRs

    International Nuclear Information System (INIS)

    Kulkarni, A.K.; Balaraman, V.; Purandare, H.D.

    1997-01-01

    A new higher order method is proposed for obtaining flux map using single set of expansion mode. In this procedure, one can make use of the difference between predicted value of detector reading and their actual values for determining the strength of local fluxes around detector site. The local fluxes are arising due to constant perturbation changes (both extrinsic and intrinsic) taking place in the reactor. (author)

  14. Corrosion of research reactor aluminium clad spent fuel in water

    International Nuclear Information System (INIS)

    2009-12-01

    A large variety of research reactor spent fuel with different fuel meats, different geometries and different enrichments in 235 U are presently stored underwater in basins located around the world. More than 90% of these fuels are clad in aluminium or aluminium based alloys that are notoriously susceptible to corrosion in water of less than optimum quality. Some fuel is stored in the reactor pools themselves, some in auxiliary pools (or basins) close to the reactor and some stored at away-from-reactor pools. Since the early 1990s, when corrosion induced degradation of the fuel cladding was observed in many of the pools, corrosion of research reactor aluminium clad spent nuclear fuel stored in light water filled basins has become a major concern, and programmes were implemented at the sites to improve fuel storage conditions. The IAEA has since then established a number of programmatic activities to address corrosion of research reactor aluminium clad spent nuclear fuel in water. Of special relevance was the Coordinated Research Project (CRP) on Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase I) initiated in 1996, whose results were published in IAEA Technical Reports Series No. 418. At the end of this CRP it was considered necessary that a continuation of the CRP should concentrate on fuel storage basins that had demonstrated significant corrosion problems and would therefore provide additional insight into the fundamentals of localized corrosion of aluminium. As a consequence, the IAEA started a new CRP entitled Corrosion of Research Reactor Aluminium Clad Spent Fuel in Water (Phase II), to carry out more comprehensive research in some specific areas of corrosion of aluminium clad spent nuclear fuel in water. In addition to this CRP, one of the activities under IAEA's Technical Cooperation Regional Project for Latin America Management of Spent Fuel from Research Reactors (2001-2006) was corrosion monitoring and surveillance of research

  15. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  16. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Belblidia, L.A.; Chevrier, A.

    1994-01-01

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  17. Sea water desalination using nuclear reactors

    International Nuclear Information System (INIS)

    Nisan, S.

    2003-01-01

    The paper first underlines the water shortage problem today and in the years to come when, around the time horizon 2020, two-thirds of the total world population would be without access to potable water. Desalination of sea-water (and, to a limited extent, that of brackish water) is shown to be an attractive solution. In this context, sea-water desalination by nuclear energy appears to be not only technically feasible and safe but also economically very attractive and a sustainable solution. Thus, compared to conventional fossil energy based sources, desalination costs by nuclear options could be 30 to 60% lower. The nuclear options are therefore expected to satisfy the fundamental water needs and electricity demands of human beings without in any way producing large amounts of greenhouse gases which any desalination strategy, based on the employment of fossil fuels, cannot fail to avoid. (author)

  18. Operation management of the prototype heavy water reactor 'Fugen'

    International Nuclear Information System (INIS)

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi

    1983-01-01

    The advanced thermal reactor Fugen power station has continued almost smooth operation since it began the full scale operation as the first homemade power reactor in Japan in March, 1979. In the initial period of operation, some troubles were experienced, but now, it can be said that the operational techniques of heavy water-moderated, boiling light water-cooled, pressure tube type reactors have been established, through the improvement of the operational method and equipment, and the operational experience. Also, the verification of the operational ability, maintainability, reliability and safety of this new type reactor, that is the mission of the prototype reactor, achieved steadily the good results. Hereafter, the verification of operational performance is the main objective because it is required for the design, construction and operation of the demonstration reactor. The organization for the operation management and operation, the communication at the time of the abnormality, the operation of the plant, that is, start up, stop and the operation at the rated output, the works during plant stoppage, the operation at the time of the plant abnormality, the operation of waste treatment facility and others, the improvement of the operational method, and the education and training of operators are reported. (Kako, I.)

  19. The design features of integrated modular water reactor (IMR)

    International Nuclear Information System (INIS)

    Kanagawa, T.; Goto, M.; Usui, S.; Suzuta, T.; Serizawa, A.; Kunugi, T.; Yamauchi, T.; Itoh, G.; Matsumura, T.

    2004-01-01

    Small-to-medium-sized (300-600 MWe) reactors are required for the electric power market in the near future (2010-2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and small-scale design is needed. From such point of view, Integrated Modular Water Reactor (IMR), whose electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of the reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid systems, and Instrumentation and Control systems of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented. (authors)

  20. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

    Purpose: To increase the production amount of fissionable plutonium by increasing the burnup degree of blanket fuels in a water cooled reactor with blanket. Method: Incore insertion assemblies comprising water elimination rods, fertile material rods or burnable poison rods are inserted to those fuel assemblies at the central portion of the reactor core that are situated at the positions not inserted with control rods in the earlier half of the operation cycle, while the incore reactor insertion assemblies are withdrawn at the latter half of the operation cycle of a nuclear reactor. As a result, it is possible to increase the power share of the blanket fuels and increase the fuel burnup degree to thereby increase the production amount of fissionable plutonium. Furthermore, at the initial stage of the cycle, the excess reactivity of the reactor can be suppressed to decrease the reactivity control share on the control rod. At the final stage of the cycle, the excess reactivity of the reactor core can be increased to improve the cycle life. (Kamimura, M.)

  1. Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

    International Nuclear Information System (INIS)

    Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

    1981-09-01

    The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized

  2. Clinch river breeder reactor plant steam generator water quality

    International Nuclear Information System (INIS)

    Van Hoesen, D.; Lowe, P.A.

    1975-01-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

  3. Clinch river breeder reactor plant steam generator water quality

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, D; Lowe, P A

    1975-07-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

  4. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

  5. Experiences in stability testing of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Otaduy, P.J.

    1986-01-01

    The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

  6. Reactivity margins in heavy water moderated production reactors

    International Nuclear Information System (INIS)

    Benton, F.D.

    1981-11-01

    The design of the reactor core and components of the heavy water moderated reactors at the Savannah River Plant (SFP) can be varied to produce a number of isotopes. For the past decade, the predominant reactor core design has been the enriched-depleted lattice. In this lattice, fuel assemblies of highly enriched uranium and target assemblies of depleted uranium, which produce plutonium, occupy alternate lattice positions. This heterogeneous lattice arrangement and a nonuniform control rod distribution result in a reactor core that requires sophisticated calculational methods for accurate reactivity margin and power distribution predictions. For maximum accuracy, techniques must exist to provide a base of observed data for the calculations. Frequent enriched-depleted lattice design changes are required as product demands vary. These changes provided incentive for the development of techniques to combine the results of calculations and observed reactivity data to accurately and conveniently monitor reactivity margins during operation

  7. Boiling water reactor modeling capabilities of MMS-02

    International Nuclear Information System (INIS)

    May, R.S.; Abdollahian, D.A.; Elias, E.; Shak, D.P.

    1987-01-01

    During the development period for the Modular Modeling System (MMS) library modules, the Boiling Water Reactor (BWR) has been the last major component to be addressed. The BWRX module includes models of the reactor core, reactor vessel, and recirculation loop. A pre-release version was made available for utility use in September 1983. Since that time a number of changes have been incorporated in BWRX to (1) improve running time for most transient events of interest, (2) extend its capability to include certain events of interest in reactor safety analysis, and (3) incorporate a variety of improvements to the module interfaces and user input formats. The purposes of this study were to briefly review the module structure and physical models, to point the differences between the MMS-02 BWRX module and the BWRX version previously available in the TESTREV1 library, to provide guidelines for choosing among the various user options, and to present some representative results

  8. Safety research for evolutionary light water reactors

    International Nuclear Information System (INIS)

    Cacuci, D.G.

    1996-01-01

    The development of nuclear energy has been characterized by a continuous evolution of the technological and philosophical underpinnings of reactor safety to enable operation of the plant without causing harm to either the plant operators or the public. Currently, the safety of a nuclear plant is assured through the combined use of procedures and engineered safety features together with a system of multiple protective barriers against the release of radioactivity. This approach is embodied in the concept of Design-Basis Accidents (DBA), which requires the designers to demonstrate that all credible accidents have been identified and that all safety equipment and procedures perform their functions extremely reliably. Particularly important functions are the automatic protection to shut the reactor down and to remove the decay heat while ensuring the integrity of the containment structure. Within the DBA concept, the so-called severe accidents were conveniently defined to be those accidents that lie beyond the DBA envelope; hence, they did not form part of the safety case. (author)

  9. Self-Sustaining Thorium Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenspan, Ehud [Univ. of California, Berkeley, CA (United States); Gorman, Phillip M. [Univ. of California, Berkeley, CA (United States); Bogetic, Sandra [Univ. of California, Berkeley, CA (United States); Seifried, Jeffrey E. [Univ. of California, Berkeley, CA (United States); Zhang, Guanheng [Univ. of California, Berkeley, CA (United States); Varela, Christopher R. [Univ. of California, Berkeley, CA (United States); Fratoni, Massimiliano [Univ. of California, Berkeley, CA (United States); Vijic, Jasmina J. [Univ. of California, Berkeley, CA (United States); Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Hall, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Ward, Andrew [Univ. of Michigan, Ann Arbor, MI (United States); Jarrett, Michael [Univ. of Michigan, Ann Arbor, MI (United States); Wysocki, Aaron [Univ. of Michigan, Ann Arbor, MI (United States); Xu, Yunlin [Univ. of Michigan, Ann Arbor, MI (United States); Kazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Shirvan, Koroush [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mieloszyk, Alexander [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Todosow, Michael [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, Nicolas [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, Lap [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-03-15

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  10. Self-Sustaining Thorium Boiling Water Reactors

    International Nuclear Information System (INIS)

    Greenspan, Ehud; Gorman, Phillip M.; Bogetic, Sandra; Seifried, Jeffrey E.; Zhang, Guanheng; Varela, Christopher R.; Fratoni, Massimiliano; Vijic, Jasmina J.; Downar, Thomas; Hall, Andrew; Ward, Andrew; Jarrett, Michael; Wysocki, Aaron; Xu, Yunlin; Kazimi, Mujid; Shirvan, Koroush; Mieloszyk, Alexander; Todosow, Michael; Brown, Nicolas; Cheng, Lap

    2015-01-01

    The primary objectives of this project are to: Perform a pre-conceptual design of a core for an alternative to the Hitachi proposed fuel-self- sustaining RBWR-AC, to be referred to as a RBWR-Th. The use of thorium fuel is expected to assure negative void coefficient of reactivity (versus positive of the RBWR-AC) and improve reactor safety; Perform a pre-conceptual design of an alternative core to the Hitachi proposed LWR TRU transmuting RBWR-TB2, to be referred to as the RBWR-TR. In addition to improved safety, use of thorium for the fertile fuel is expected to improve the TRU transmutation effectiveness; Compare the RBWR-Th and RBWR-TR performance against that of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; and, Perform a viability assessment of the thorium-based RBWR design concepts to be identified along with their associated fuel cycle, a technology gap analysis, and a technology development roadmap. A description of the work performed and of the results obtained is provided in this Overview Report and, in more detail, in the Attachments. The major findings of the study are summarized.

  11. Safety research for evolutionary light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cacuci, D G [Karlsruhe Univ. (T.H.) (Germany). Universitaetsbibliothek

    1996-12-01

    The development of nuclear energy has been characterized by a continuous evolution of the technological and philosophical underpinnings of reactor safety to enable operation of the plant without causing harm to either the plant operators or the public. Currently, the safety of a nuclear plant is assured through the combined use of procedures and engineered safety features together with a system of multiple protective barriers against the release of radioactivity. This approach is embodied in the concept of Design-Basis Accidents (DBA), which requires the designers to demonstrate that all credible accidents have been identified and that all safety equipment and procedures perform their functions extremely reliably. Particularly important functions are the automatic protection to shut the reactor down and to remove the decay heat while ensuring the integrity of the containment structure. Within the DBA concept, the so-called severe accidents were conveniently defined to be those accidents that lie beyond the DBA envelope; hence, they did not form part of the safety case. (author).

  12. Water chemistry management of research reactor in JAERI

    Energy Technology Data Exchange (ETDEWEB)

    Yoshijima, Tetsuo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-10-01

    The JRR-3M cooling system consists of four systems, namely; (1) primary cooling system, (2) heavy water cooling system, (3) helium system and (4) secondary cooling system. The heavy water is used for reflector and pressurized with helium gas. Water chemistry management of the JRR-3M cooling systems is one of the important subject for the safety operation. The main objects are to prevent the corrosion of cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the generation of radioactive waste. All measured values were within the limits of specifications and JRR-3M reactor was operated with safety in 1996. Spent fuels of JRR-3M reactor are stored in the spent fuel pool. This pool water has been analyzed to prevent corrosion of aluminum cladding of spent fuels. Water chemistry of spent fuel pool water is applied to the prevention of corrosion of aluminum alloys including fuel cladding. The JRR-2 reactor was eternally stopped in December 1996 and is now under decommissioning. The JRR-2 reactor is composed of heavy water tank, fuel guide tube and horizontal experimental hole. These are constructed of aluminum alloy and biological shield and upper shield are constructed of concrete. Three types of corrosion of aluminum alloy were observed in the JRR-2. The Alkaline corrosion of aluminum tube occurred in 1972 because of the mechanical damage of the aluminum fuel guide tube which is used for fuel handling. Modification of the reactor top shield was started in 1974 and completed in 1975. (author)

  13. 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

  14. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Power control device for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Matsushima, Hidesuke; Masuda, Hiroyuki.

    1978-01-01

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

  16. Development of next-generation light water reactor

    International Nuclear Information System (INIS)

    Ishibashi, Fumihiko; Yasuoka, Makoto

    2010-01-01

    The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

  17. Heavy-Water Power Reactors. Proceedings Of A Symposium

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1968-04-15

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  18. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  19. Heavy-Water Power Reactors. Proceedings Of A Symposium

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held in Vienna, 11-15 September 1967. The timeliness of the meeting was underlined by the large gathering of over 225 participants from 28 countries and three international organizations. Contents: Experience with heavy-water power and experimental reactors and projects (14 papers); New and advanced power reactor designs and concepts (8 papers); Development programmes and thorium cycle (9 papers); Economics and prospects of heavy-water power reactors (7 papers); Physics and fuel management (8 papers); Fuels (5 papers); Safety, control and engineering (6 papers); Panel discussion. Except for one Russian paper, which is published in English, each paper is in its original language (49 English and 8 French) and is preceded by an abstract in English with a second one in the original language if this is not English. Discussions are in English. (author)

  20. Balancing passive and active systems for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    Fil, N.S.; Allen, P.J.; Kirmse, R.E.; Kurihara, M.; Oh, S.J.; Sinha, R.K.

    1999-01-01

    Advanced concepts of the water-cooled reactors are intended to improve safety, economics and public perception of nuclear power. The potential inclusion of new passive means in addition or instead of traditional active systems is being considered by nuclear plant designers to reach these goals. With respect to plant safety, application of the passive means is mainly intended to simplify the safety systems and to improve their reliability, to mitigate the effect of human errors and equipment malfunction. However, some clear drawbacks and the limited experience and testing of passive systems may raise additional questions that have to be addressed in the design process for each advanced reactor. Therefore the plant designer should find a reasonable balance of active and passive means to effectively use their advantages and compensate their drawbacks. Some considerations that have to be taken into account when balancing active/passive means in advanced water-cooled reactors are discussed in this paper. (author)

  1. High conversion ratio plutonium recycle in pressurized water reactors

    International Nuclear Information System (INIS)

    Edlund, M.C.

    1975-01-01

    The use of Pu light water reactors in such a way as to minimise the depletion of Pu needed for future use, and therefore to reduce projected demands for U ore and U enrichment is envisaged. Fuel utilisation in PWRs could be improved by tightly-packed fuel rod lattices with conversion ratios of 0.8 to 0.9 compared with ratios of about 0.5 in Pu recycle designs using fuel to water volume ratios of currently operating PWRs. A conceptual design for the Babcock and Wilcox Company reactors now in operation is presented and for illustrative purposes thermalhydraulic design considerations and the reactor physics are described. Principle considerations in the mechanical design of the fuel assemblies are the effect of hydraulic forces, thermal expansion, and fission gas release. The impact of high conversion ratio plutionium recycle in separative work and natural U requirements for PWRs likely to be in operation by 1985 are examined. (U.K.)

  2. Aerosol behavior and light water reactor source terms

    International Nuclear Information System (INIS)

    Abbey, F.; Schikarski, W.O.

    1988-01-01

    The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered

  3. Modelling chemical behavior of water reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Ball, R G.J.; Hanshaw, J; Mason, P K; Mignanelli, M A [AEA Technology, Harwell (United Kingdom)

    1997-08-01

    For many applications, large computer codes have been developed which use correlation`s, simplifications and approximations in order to describe the complex situations which may occur during the operation of nuclear power plant or during fault scenarios. However, it is important to have a firm physical basis for simplifications and approximations in such codes and, therefore, there has been an emphasis on modelling the behaviour of materials and processes on a more detailed or fundamental basis. The application of fundamental modelling techniques to simulated various chemical phenomena in thermal reactor fuel systems are described in this paper. These methods include thermochemical modelling, kinetic and mass transfer modelling and atomistic simulation and examples of each approach are presented. In each of these applications a summary of the methods are discussed together with the assessment process adopted to provide the fundamental parameters which form the basis of the calculation. (author). 25 refs, 9 figs, 2 tabs.

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

    International Nuclear Information System (INIS)

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

    1963-01-01

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

  5. 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

  6. A review of boiling water reactor water chemistry: Science, technology, and performance

    International Nuclear Information System (INIS)

    Fox, M.J.

    1989-02-01

    Boiling water reactor (BWR) water chemistry (science, technology, and performance) has been reviewed with an emphasis on the relationships between BWR water quality and corrosion fuel performance, and radiation buildup. A comparison of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.56, the Boiling Water Reactor Owners Group (BWROG) Water Chemistry Guidelines, and Plant Technical Specifications showed that the BWROG Guidelines are more stringent than the NRC Regulatory Guide, which is almost identical to Plant Technical Specifications. Plant performance with respect to BWR water chemistry has shown dramatic improvements in recent years. Up until 1979 BWRs experienced an average of 3.0 water chemistry incidents per reactor-year. Since 1979 the water chemistry technical specifications have been violated an average of only 0.2 times per reactor-year, with the most recent data from 1986-1987 showing only 0.05 violations per reactor-year. The data clearly demonstrate the industry-wide commitment to improving water quality in BWRs. In addition to improving water quality, domestic BWRs are beginning to switch to hydrogen water chemistry (HWC), a remedy for intergranular stress corrosion cracking. Three domestic BWRs are presently operating on HWC, and fourteen more have either performed HWC mini tests or are in various stages of HWC implementation. This report includes a detailed review of HWC science and technology as well as areas in which further research on BWR chemistry may be needed. 43 refs., 30 figs., 8 tabs

  7. Spectral shift rod for the boiling water reactor

    International Nuclear Information System (INIS)

    Yokomizo, O.; Kashiwai, S.; Nishida, K.; Orii, A.; Yamashita, J.; Mochida, T.

    1993-01-01

    A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained. (orig.)

  8. European community light water reactor safety research projects. Experimental issue

    International Nuclear Information System (INIS)

    1975-01-01

    Research programs on light water reactor safety currently carried out in the European Community are presented. They cover: accident conditions (LOCA, ECCS, core meltdown, external influences, etc...), fault and accident prevention and means of mitigation, normal operation conditions, on and off site implications and equipment under severe accident conditions, and miscellaneous subjects

  9. Thermophysical properties of materials for water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-06-01

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

  10. Thermophysical properties of materials for water cooled reactors

    International Nuclear Information System (INIS)

    1997-06-01

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

  11. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  12. Impingement studies at the 100-N reactor water intake

    International Nuclear Information System (INIS)

    Page, T.L.; Neitzel, D.A.; Gray, R.H.

    1977-09-01

    Fish impingement and traveling screen passage were studied at the 100-N reactor water intake structure, Columbia River mile 380, from late April to August 1977. Species and numbers of fish affected were determined and compared to those at the adjacent Hanford Generating Project (HGP). Fish protection procedures previously developed for HGP were evaluated for application at 100-N

  13. Gravity Scaling of a Power Reactor Water Shield

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined

  14. Loose parts monitoring in light water reactor cooling systems

    International Nuclear Information System (INIS)

    Santos, A.; Alma, B.J.

    1982-01-01

    The work related to loose monitoring system for light water reactor, developed at GRS - Munique, are described. The basic problems due to the exact localization and detection of the loose part as well the research activities and development necessary aiming to obtain the best techniques in this field. (E.G.) [pt

  15. The manufacture of plutonium fuels for light water reactors

    International Nuclear Information System (INIS)

    Lebastard, G.

    1985-01-01

    This paper describes the agreement concluded between COGEMA and BELGONUCLEAIRE, reflected in the creation of the COMMOX group which has been made reponsible for promoting and marketing plutonium fuel rods for light water reactors. One then analyses the main aspects of manufacturing this type of fuel and the resources deployed. Finally one indicates the sales prospects scheduled to meet requirements (MELOX plant) [fr

  16. Method of controlling the water quality in nuclear reactors

    International Nuclear Information System (INIS)

    Ibe, Hidefumi.

    1985-01-01

    Purpose: To obtain a simple and reliable water quality calculation system and water quality control method based thereon for the entire primary coolant circuits in BWR type reactors. Method: In a method of controlling the water quality of the reactor water by injecting hydrogen into the primary coolant circuits of a nuclear reactor, by utilizing a first linear relationship established between the concentration of oxygen and hydrogen in the main steam system and the concentration of radiolysis products in the reactor core and separators and mixing plenum portions, each of the above-mentioned concentrations is calculated from the concentrations for hydrogen or oxygen. Further, by utilizing the first linear relationship established between the concentrations for the oxygen and hydrogen in the recycling system and the concentration of the radiolysis products in the system from the downcomer to the lower plenum portion, the above-mentioned concentration is calculated from the concentration for oxygen and hydrogen. Then, the hydrogen injection rate into the primary coolant system is determined such that the calculated value takes an aimed value. (Ikeda, J.)

  17. Tritium formation and elimination in light-water reactors

    International Nuclear Information System (INIS)

    Dolle, L.; Briec, M.; Miquel, P.

    1976-01-01

    Light-water reactors have a tritium balance which should be considered from both the working constraint and environmental pollution aspects. The formation of tritium in the primary circuit and in the fuel, the elimination and enrichment processes are considered [fr

  18. Design features to facilitate IAEA safeguards at light water reactors

    International Nuclear Information System (INIS)

    Pasternak, T.; Glancy, J.; Goldman, L.; Swartz, J.

    1981-01-01

    Several studies have been performed recently to identify and analyze light water reactor (LWR) features that, if incorporated into the facility design, would facilitate the implementation of International Atomic Energy Agency (IAEA) safeguards. This paper presents results and conclusions of these studies. 2 refs

  19. 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.

  20. 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

  1. Framework for applying RI-ISI methodology for Indian PHWRs

    International Nuclear Information System (INIS)

    Vinod, Gopika; Saraf, R.K.; Ghosh, A.K.; Kushwaha, H.S.

    2006-01-01

    Risk Informed In-Service Inspection (RI-ISI) aims at categorizing the components for In-Service inspection based on their contribution to Risk. For defining the contribution of risk from components, their failure probabilities and its subsequent effect on Core Damage Frequency (CDF) needs to be evaluated using Probabilistic Safety Assessment methodology. During the last several years, both the U.S. Nuclear Regulatory Commission (NRC) and the nuclear industry have recognized that Probabilistic Safety Assessment (PSA) has evolved to be more useful in supplementing traditional engineering approaches in reactor regulation. The paper highlights the various stages involved in applying RI-ISI and then compares the findings with existing ISI practices. (author)

  2. 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

  3. Measuring device for water quality at reactor bottom

    Energy Technology Data Exchange (ETDEWEB)

    Urata, Hidehiro; Takagi, Jun-ichi

    1995-10-27

    The present invention concerns measurement for water quality at the bottom of a reactor of a BWR type plant, in which reactor water is sampled and analyzed in a state approximate to conditions in a pressure vessel. Based on the result, hydrogen injection amount is controlled during hydrogen injection operation. Namely, a monitor for water quality is disposed to a sampling line in communication with the bottom of a pressure vessel. A water quality monitor is disposed to a drain sampling line in communication with the bottom of the pressure vessel. A corrosion potentiometer is disposed to the pressure sampling line or the drain sampling line. A dissolved oxygen measuring device is disposed to the pressure vessel sampling line or the drain sampling line. With such a constitution, the reactor water can be sampled and analyzed in a state approximate to the conditions in the pressure vessel. In addition, signals from the water quality monitor are inputted to a hydrogen injection amount control device. As a result, the amount of hydrogen injected to primary coolants can be controlled in a state approximate to the conditions in the pressure vessel. (I.S.).

  4. Measuring device for water quality at reactor bottom

    International Nuclear Information System (INIS)

    Urata, Hidehiro; Takagi, Jun-ichi.

    1995-01-01

    The present invention concerns measurement for water quality at the bottom of a reactor of a BWR type plant, in which reactor water is sampled and analyzed in a state approximate to conditions in a pressure vessel. Based on the result, hydrogen injection amount is controlled during hydrogen injection operation. Namely, a monitor for water quality is disposed to a sampling line in communication with the bottom of a pressure vessel. A water quality monitor is disposed to a drain sampling line in communication with the bottom of the pressure vessel. A corrosion potentiometer is disposed to the pressure sampling line or the drain sampling line. A dissolved oxygen measuring device is disposed to the pressure vessel sampling line or the drain sampling line. With such a constitution, the reactor water can be sampled and analyzed in a state approximate to the conditions in the pressure vessel. In addition, signals from the water quality monitor are inputted to a hydrogen injection amount control device. As a result, the amount of hydrogen injected to primary coolants can be controlled in a state approximate to the conditions in the pressure vessel. (I.S.)

  5. Aging management of major light water reactor components

    International Nuclear Information System (INIS)

    Shah, V.N.; Sinha, U.P.; Ware, A.G.

    1992-01-01

    Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

  6. Mathematical modeling of water radiolysis in the Syrian MNSR reactor

    International Nuclear Information System (INIS)

    Soukieh, M.

    2009-11-01

    Because it is difficult to measure the concentration of the radiolytic species in reactors under operating conduction, they must be estimated by computer simulation techniques. This study discusses the mathematical modeling of water radiolysis modeling of the MNSR nuclear reactor cooling water. The mathematical model comprising of 13 differential equations describe 55 chemical reactions of radiolytic species e - a q H + , OH - , H, H 2 , OH, HO 2 , O 2 , HO - 2 , O - , O - 2 , O - 3 . The mathematical model have been tested and it shows a good agreement of the computed values in this work with the results cited in references [1,18] in case of only γray irradiation of pure water with dose rate of 1.18x10 19 eV/L s. The neutron fluxes and dose rates at the interface of cladding-water for the different fuel rings in the MNSR core are determined using MCNP-4C code. In addition, the time dependent of the radiolytic specie concentrations were estimated for max. and min. dose rates and at temperature of 20 degree centigrade in the MNSR. The radiolytic specie concentrations reach the steady sate after about 200-400 s. The radiolytic specie concentrations order of H 2 , O 2 , H 2 O 2 were about ppb. Also this study shows the possibility of suppressed the water radiolysis reactions by adding hydrogen to the MNSR reactor cooling water. (author)

  7. Mechanical design of a light water breeder reactor

    International Nuclear Information System (INIS)

    Fauth, W.L. Jr.; Jones, D.S.; Kolsun, G.J.; Erbes, J.G.; Brennan, J.J.; Weissburg, J.A.; Sharbaugh, J.E.

    1976-01-01

    In a light water reactor system using the thorium-232--uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements. 4 claims, 24 drawing figures

  8. Environmentally assisted cracking of light-water reactor materials

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

    1996-02-01

    Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used

  9. Development of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Yamashita, Jun-ichi; Mochida, Takaaki; Uchikawa, Sadao.

    1988-01-01

    It is expected that the period of LWRs being the main source of electric power supply becomes long, therefore, the development of next generation LWRs placing emphasis on the effective utilization of uranium resources and the improvement of economical efficiency is necessary. In this paper, as the next generation BWRs subsequent to ABWRs, the concept of the core of high conversion type BWRs is reported, in which emphasis is placed on the saving of natural uranium resources by raising the rate of conversion to plutonium. This core is that of realizing the high rate of conversion by utilizing the void in the core, which is the feature of BWRs, and the case of making the change of the core structure relatively small by using cross type control rods and the case of changing the core structure for further heightening the rate of conversion and making control rods into cluster type are described. In order to meet the demand like this, Hitachi Ltd. has engaged in the development of the concept of the core of next generation LWRs. In the high conversion type BWRs, there is not large change in the reactor system and turbine system from the current BWRs. The features and the concept of the core of high conversion type BWRs are described. (Kako, I.)

  10. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2001-06-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts, as well as for current nuclear applications Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for the coupling of core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for the purpose. The present volume describes the specification of such a benchmark. The transient addressed is a turbine trip (TT) in a BWR involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the plant make the present benchmark very valuable. The data used are from events at the Peach Bottom 2 reactor (a GE-designed BWR/4). (authors)

  11. Contaminants in light water reactor coolants

    International Nuclear Information System (INIS)

    Michael, I.; Bechtold, G.

    1975-01-01

    At a lower oxygen content of the pressurized water a reduced metal loss by about 10% was detected. The state of oxidation for incoloy resulting from surface examination was 2,3 +- 0,3 which corresponds to Fe 3 O 4 and a smaller fraction of iron hydroxide. (orig.) [de

  12. Steam generator for pressurized-water reactors

    International Nuclear Information System (INIS)

    Michel, E.

    1971-01-01

    In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

  13. Accident analysis of heavy water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  14. Accident analysis of heavy water cooled thorium breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-16

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

  15. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    International Nuclear Information System (INIS)

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

  16. The Consortium for Advanced Simulation of Light Water Reactors

    International Nuclear Information System (INIS)

    Szilard, Ronaldo; Zhang, Hongbin; Kothe, Douglas; Turinsky, Paul

    2011-01-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  17. Fuel cycle options for light water reactors and heavy water reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1999-11-01

    In the second half of the 20th century nuclear power has evolved from the research and development environment to an industry that supplies 16% of the world's electricity. By the end of 1997, over 8500 reactor-years of operating experience had been accumulated. Global environmental change, and the continuing increase in global energy supply required to provide increasing populations with an improving standard of living, make the contribution from nuclear energy even more important for the next century. For nuclear power to achieve its full potential and make its needed contribution, it must be safe, economical, reliable and sustainable. All of these factors can be enhanced by judicious choice and development of advanced fuel cycle options. The Technical Committee Meeting (TCM) on Fuel Cycle Options for Light Water Reactors and Heavy Water Reactors was hosted by Atomic Energy of Canada Limited (AECL) on behalf of the Canadian Government and was jointly conducted within the frame of activities of the IAEA International Working Group on Advanced Technologies for Light Water Reactors (IWG-LWR) and the IAEA International Working Group on Advanced Technologies for Heavy Water Reactors (IWG-HWR). The TCM provided the opportunity to have in-depth discussions on important technical topics which were highlighted in the International Symposium on Nuclear Fuel Cycle and Reactor Strategies: Adjusting to New Realities, held in Vienna, 3-6 June 1997. The main results and conclusions of the TCM were presented as input for discussion at the first meeting of the IAEA newly formed International Working Group on Fuel Cycle Options

  18. Development of reactor water level sensor for extreme conditions

    Energy Technology Data Exchange (ETDEWEB)

    Miura, K; Ogasawara, T [Sukegawa Electric Co., Ltd., Hitachi, Ibaraki (Japan); Shibata, Akira; Nakamura, Jinichi; Saito, Takashi; Tsuchiya, Kunihiko [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

    2012-03-15

    In the Fukushima accident, measurement failure of water level was one of the most important factors which caused serious situation. The differential pressure type water level indicators are widely used in various place of nuclear power plant but after the accident of TMI-2, the need of other reliable method has been required. The BICOTH type and the TRICOTH type water level indicator for light water power reactors had been developed for in-pile water level indicator but currently those are not adopted to nuclear power plant. In this study, the development of new type water level indicator composed of thermocouple and heater is described. Demonstration test and characteristic evaluation of the water level indicator were performed and we had obtained satisfactory results. (author)

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

    International Nuclear Information System (INIS)

    Cleveland, J.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  1. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  2. Computational fluid dynamics simulations of light water reactor flows

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Weber, D.P.

    1999-01-01

    Advances in computational fluid dynamics (CFD), turbulence simulation, and parallel computing have made feasible the development of three-dimensional (3-D) single-phase and two-phase flow CFD codes that can simulate fluid flow and heat transfer in realistic reactor geometries with significantly reduced reliance, especially in single phase, on empirical correlations. The objective of this work was to assess the predictive power and computational efficiency of a CFD code in the analysis of a challenging single-phase light water reactor problem, as well as to identify areas where further improvements are needed

  3. Topical papers on heavy water, fuel fabrication and reactors

    International Nuclear Information System (INIS)

    1978-01-01

    A total of four papers is presented. The first contribution of the Federal Republic of Germany reviews the market situation for reactors and the relations between reactor producers and buyers as reflected in sales agreements. The second West German contribution gives a world-wide survey of fuel element production as well as of fuel and fuel element demand up to the year 2000. The Canadian paper discusses the future prospects of heavy-water production, while the Ecuador contribution deals with small and medium-sized nuclear power plants

  4. The industry/EPRI advanced light water reactor program

    International Nuclear Information System (INIS)

    Stahlkopf, K.E.; Noble, D.M.; Sugnet, W.R.; Bilan, W.J.

    1986-01-01

    For the United States nuclear power industry to remain viable, it must be prepared to meet the expected need for new generating capacity in the late 1990s with an improved reactor system. The best hope of meeting this requirement is with evolutionary changes in current LWR systems through system simplification and reevaluation of safety and operational design margins. The grid characteristics and the difficulty in raising capital for large projects indicate that smaller light water reactors (400 to 600 MWe) may play an important role the next generation

  5. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  6. Major outage trends in light water reactors. Interim report

    International Nuclear Information System (INIS)

    Burns, E.T.

    1978-04-01

    The report is a summary of the major outages which occurred in light water reactor plants during the period January 1971 through June 1977. Only those outages greater than 100 hours duration (exclusive of refueling outages) are included in the report. The trends in outages related to various reactor systems and components are presented as a function of plant age, and alternatively, calendar year. The principal contributors to major outages are ranked by their effect on the overall outage time for PWRs and BWRs. In addition, the outage history of each operating nuclear plant greater than 150 MWe is presented, along with a brief summary of those outages greater than two months duration

  7. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  8. Automated ultrasonic examination of light water reactor systems

    International Nuclear Information System (INIS)

    Walter, J.H.

    1975-01-01

    An automated ultrasonic examination system has been developed to meet the pre- and inservice inspection requirements of light water reactors. This system features remotely-controlled travelling instrument carriers, computerized collection and storage or inspection data in a manner providing real time comparison against code standards, and computer control over the positioning of the instrument carriers to provide precise location data. The system is currently being utilized in the field for a variety of reactor inspections. The principal features of the system and the recent inspection experience are discussed. (author)

  9. Transactions of the Twentieth Water Reactor Safety Information Meeting

    International Nuclear Information System (INIS)

    Weiss, A.J.

    1992-10-01

    This report contains summaries of papers on reactor safety research to be presented at the 20th Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel in Bethesda, Maryland, October 21--23, 1992. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, USNRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included

  10. Stainless steel clad for light water reactor fuels. Final report

    International Nuclear Information System (INIS)

    Rivera, J.E.; Meyer, J.E.

    1980-07-01

    Proper reactor operation and design guidelines are necessary to assure fuel integrity. The occurrence of fuel rod failures for operation in compliance with existing guidelines suggests the need for more adequate or applicable operation/design criteria. The intent of this study is to develop such criteria for light water reactor fuel rods with stainless steel clad and to indicate the nature of uncertainties in its development. The performance areas investigated herein are: long term creepdown and fuel swelling effects on clad dimensional changes and on proximity to clad failure; and short term clad failure possibilities during up-power ramps

  11. Advances in commercial heavy water reactor power stations

    International Nuclear Information System (INIS)

    Brooks, G.L.

    1987-01-01

    Generating stations employing heavy water reactors have now firmly established an enviable record for reliable, economic electricity generation. Their designers recognize, however, that further improvements are both possible and necessary to ensure that this reactor type remains attractively competitive with alternative nuclear power systems and with fossil-fuelled generation plants. This paper outlines planned development thrusts in a number of important areas, viz., capital cost reduction, advanced fuel cycles, safety, capacity factor, life extension, load following, operator aida, and personnel radiation exposure. (author)

  12. State of the art of the advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Chawla, R.

    1987-01-01

    A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

  13. The thorium fuel cycle in water-moderated reactor systems

    International Nuclear Information System (INIS)

    Critoph, E.

    1977-05-01

    Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An 'inventory' of uranium of between 1 and 2 Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium) is some two decades

  14. New fuel advanced heavy water reactors

    International Nuclear Information System (INIS)

    Notari, Carla

    1999-01-01

    A redesign of the PHWR fuel element (FE) to be used in all Argentine nuclear power plants has been proposed elsewhere. This new FE presents several characteristics aimed to an improved in-core performance and economical benefits derived from the unification of most of the fabrication processes that today constitute two different production lines: one for Embalse nuclear power plant CANDU type fuel and another for Atucha I. Atucha I and Embalse, the two operating nuclear power plants in Argentina, are PHWR of different conception. Atucha I (357 M we) is of pressure vessel type and the fuel elements are full-length assemblies (530 cm of active length) with 36 uranium rods in the cluster and a support one in the outer ring. Embalse (648 M we) is a CANDU pressure tube reactor fuelled with the well known 37 rod / 50 cm length fuel bundles, twelve of which are loaded in each channel. The more relevant changes in the proposed design are an increased subdivision of the fuel material in 52 rods and a 100 cm long bundle. The combined features give the adequate channel pressure drop. The proposed CARA design shows a superior neutronic performance than the standard PHWR fuel elements currently used in Atucha I and Embalse nuclear power plants. A variant of the CARA FE consisting in the elimination of the central four rods, leaving 48 rods and a central free space, is strongly recommended because it saves materials (less uranium, less sheaths) with no loss of burnup. The central D 2 O zone allows a better utilization of the inner rods and compensates the diminished uranium loading. In Embalse no differences in core physics are expected except the beneficial decrease in linear power density. In Atucha I besides the lower power density, a higher exit burnup appears as a consequence of the higher uranium inventory. The exit burnup figures have been calculated with cell and reactor models and the result is that similar fuel management schemes as the proposed for Atucha I for the

  15. Fixed-biofilm reactors applied to waste water treatment and aquacultural water recirculating systems

    NARCIS (Netherlands)

    Bovendeur, J.

    1989-01-01

    Fixed-biofilm waste water treatment may be regarded as one of the oldest engineered biological waste water treatment methods. With the recent introduction of modern packing materials, this type of reactor has received a renewed impuls for implementation in a wide field of water treatment.

    In

  16. The light water natural uranium reactor

    International Nuclear Information System (INIS)

    Radkowsky, A.

    A new type of light water seed blanket with the seed having 20% enrichment and the blanket a special combination of elements of natural uranium and thorium, relatively close packed, but sufficient spacing for heat transfer purpose is described. The blanket would deliver approximately half the total energy for about 10,000 MWDIT, so this type of core would be just as economical or better in uranium ore consumation as present cores. (author)

  17. Steam plant for pressurized water reactors

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

  18. Study on core design for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Okubo, Tsutomu

    2002-01-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  19. Dual-purpose light water reactor supplying heat for desalination

    International Nuclear Information System (INIS)

    Waplington, G.; Fichtner, H.

    1978-01-01

    The technical as well as the economic aspects of using a large commercial light water reactor for the production of both electricity and potable water have been examined. For the basis of the study, the multistage flash distillation process was selected, in conjunction with a reactor rated at not less than 2100 MW (thermal). Combined use of a condensing and a back-pressure turbine (the latter matched to distillation plant steam requirements) represents a convenient method for supplying process heat. Overall costs can be fairly allocated to the two products using the ''power credit'' method. A sample economic evaluation indicates highly favorable water costs as compared with more conventional distillation schemes based on fossil fuel

  20. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  1. The thorium fuel cycle in water-moderated reactor systems

    International Nuclear Information System (INIS)

    Critoph, E.

    1977-01-01

    Current interest in the thorium cycle, as an alternative to the uranium cycle, for water-moderated reactors is based on two attractive aspects of its use - the extension of uranium resources, and the related lower sensitivity of energy costs to uranium price. While most of the scientific basis required is already available, some engineering demonstrations are needed to provide better economic data for rational decisions. Thorium and uranium cycles are compared with regard to reactor characteristics and technology, fuel-cycle technology, economic parameters, fuel-cycle costs, and system characteristics. There appear to be no major feasibility problems associated with the use of thorium, although development is required in the areas of fuel testing and fuel management. The use of thorium cycles implies recycling the fuel, and the major uncertainties are in the associated costs. Experience in the design and operation of fuel reprocessing and active-fabrication facilities is required to estimate costs to the accuracy needed for adequately defining the range of conditions economically favourable to thorium cycles. In heavy-water reactors (HWRs) thorium cycles having uranium requirements at equilibrium ranging from zero to a quarter of those for the natural-uranium once-through cycle appear feasible. An ''inventory'' of uranium of between 1 and 2Mg/MW(e) is required for the transition to equilibrium. The cycles with the lowest uranium requirements compete with the others only at high uranium prices. Using thorium in light-water reactors, uranium requirements can be reduced by a factor of between two and three from the once-through uranium cycle. The light-water breeder reactor, promising zero uranium requirements at equilibrium, is being developed. Larger uranium inventories are required than for the HWRs. The lead time, from a decision to use thorium to significant impact on uranium utilization (compared to uranium cycle, recycling plutonium), is some two decades

  2. Materials Degradation in Light Water Reactors: Life After 60,

    International Nuclear Information System (INIS)

    Busby, Jeremy T; Nanstad, Randy K; Stoller, Roger E; Feng, Zhili; Naus, Dan J

    2008-01-01

    Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase

  3. Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    De, S.K.; Karhadkar, C.G.

    2017-01-01

    An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

  4. Safety aspects of designs for future light water reactors (evolutionary reactors)

    International Nuclear Information System (INIS)

    1993-07-01

    The main purpose of this document is to describe the major innovations of proposed designs of future light water reactors, to describe specific safety characteristics and safety analysis methodologies, and to give a general overview of the most important safety aspects related to future reactors. The reactors considered in this report are limited to those intended for fixed station electrical power production, excluding most revolutionary concepts. More in depth discussion is devoted to those designs that are in a more advanced state of completion and have been more extensively described and analysed in the open literature. Other designs will be briefly described, as evidence of the large spectrum of new proposals. Some designs are similar; others implement unique features and require specific discussion (not all aspects of designs with unique features are fully discussed in this document). 131 refs, 22 figs

  5. Dynamic operator actions analysis for inherently safe fast reactors and light water reactors

    International Nuclear Information System (INIS)

    Ho, V.; Apostolakis, G.

    1988-01-01

    A comparative dynamic human actions analysis of inherently safe fast reactors (ISFRs) and light water reactors (LWRs) in terms of systems response and estimated human error rates is presented. Brief overviews of the ISFR and LWR systems are given to illustrate the design differences. Key operator actions required by the ISFR reactor shutdown and decay heat removal systems are identified and are compared with those of the LWR. It is observed that, because of the passive nature of the ISFR safety-related systems, a large time window is available for operator actions during transient events. Furthermore, these actions are fewer in number, are less complex, and have lower error rates and less severe consequences than those of the LWRs. We expect the ISFR operator errors' contribution to risk is smaller (at least in the context of the existing human reliability models) than that of the LWRs. (author)

  6. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Solomon, K.A.

    1979-07-01

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  7. Role of water lubricated bearings in Candu reactors

    International Nuclear Information System (INIS)

    Kumar, Ashok N.

    1999-01-01

    During the twentieth century a great emphasis was placed in understanding and defining the operating regime of oil and grease lubricated components. Major advances have been made through elastohydrodynamic lubrication theory in the quantifying the design life of heavily loaded components such as rolling element bearings and gears. Detailed guidelines for the design of oil and grease lubricated components are widely available and are being applied to the successful design of these components. However similar guidelines for water lubricated components are either not available or not well documented. It is often forgotten that the water was used as a lubricant in several components as far back as 1884 B.C. During the twentieth century the water lubricated components continued to play a major role in some high technology industries such as in the power generation plants. In CANDU nuclear reactors water lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication of several critical components always occupied a pride place and in most cases the only practical mode of lubrication. This paper presents some examples of the major water lubricated components in a CANDU reactors. Major part of the paper is focused on presenting an example of successful operating history of water lubricated bearings used in the HT pumps are presented. Both types of bearings have been qualified by tests for operation under normal as well as under more severe postulated condition of loss-of-coolant-accident (LOCA). These bearings have been designed to operate for the 30 years in the existing CANDU 6 (600 MW) reactors. However for the next generation of CANDU 6 reactors which go into service in the year 2003, the HT pump bearing life has been extended to 40 years. (author)

  8. European supercritical water cooled reactor (HPLWR Phase 2 project)

    International Nuclear Information System (INIS)

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

    2010-01-01

    The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small, housed fuel assemblies with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

  9. Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

    International Nuclear Information System (INIS)

    Permana, S.; Takaki, N.; Sekimoto, H.

    2007-01-01

    Thorium fuel cycle can maintain the sustainable system of the reactor for self sustaining system for future sustainable development in the world. Some characteristics of thorium cycle show some advantages in relation to higher breeding capability, higher performance of burn-up and more proliferation resistant. Several investigations was performed to improve the breeding capability which is essential for maintaining the fissile sustainability during reactor operation in thermal reactor such as Shippingport reactor and molten salt breeder reactor (MSBR) project. The preliminary study of breeding capability on water cooled thorium reactor has been investigated for various power output. The iterative calculation system is employed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000. In this calculation, 1238 fission products and 129 heavy nuclides are employed. In the cell calculation, 26 heavy metals and 66 fission products and 1 pseudo FP are employed. The employed nuclear data library was JENDL 3.2. The reactor is fueled by 2 33U-Th Oxide and it has used the light water coolant as moderator. Some characteristics such as conversion ratio and void reactivity coefficient performances are evaluated for the systems. The moderator to fuel ratio (MFR) values and average burnups are studied for survey parameter. The parametric survey for different power outputs are employed from 10 MWt to 3000 MWt for evaluating the some characteristics of core size and leakage effects to the spectra profile, required enrichment, breeding capability, fissile inventory condition, and void reactivity coefficient. Different power outputs are employed in order to evaluate its effect to the required enrichment for criticality, breeding capability, void reactivity and fissile inventory accumulation. The obtained value of the conversion ratios is evaluated by using the equilibrium atom composition. The conversion ratio is employed based on the

  10. Boiling water reactor liquid radioactive waste processing system

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

  11. COMSORS: A light water reactor chemical core catcher

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W.

    1997-01-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B 2 O 3 ) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation

  12. Comparative economics of the breeder and light water reactor

    International Nuclear Information System (INIS)

    Chow, B.G.

    1980-01-01

    The issue of breeder timing is studied in this article via a breakeven analysis in which the key driving variables are conveniently segregated into two groups, with uranium price providing the linkage. In one group, the technical and cost characteristics of reactors and fuel cycles determine the uranium breakeven price. In the other group, nuclear demand projections and the uranium supply schedule determine the time paths of uranium price for a given composition of reactor types. The author finds that, even if proliferation risk is ignored, the breeder is not economically competitive with a 30%-improved once-through light water reactor before the year 2030 in the USA and in the world outside communist areas as a whole in 90% of the cases examined. In the exceptional cases, the penalty of delaying commercial breeder introduction to 2030 is small and well within the noise level of long-term energy planning. (author)

  13. Standard Guide for Benchmark Testing of Light Water Reactor Calculations

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This guide covers general approaches for benchmarking neutron transport calculations in light water reactor systems. A companion guide (Guide E2005) covers use of benchmark fields for testing neutron transport calculations and cross sections in well controlled environments. This guide covers experimental benchmarking of neutron fluence calculations (or calculations of other exposure parameters such as dpa) in more complex geometries relevant to reactor surveillance. Particular sections of the guide discuss: the use of well-characterized benchmark neutron fields to provide an indication of the accuracy of the calculational methods and nuclear data when applied to typical cases; and the use of plant specific measurements to indicate bias in individual plant calculations. Use of these two benchmark techniques will serve to limit plant-specific calculational uncertainty, and, when combined with analytical uncertainty estimates for the calculations, will provide uncertainty estimates for reactor fluences with ...

  14. Age-related degradation of boiling water reactor vessel internals

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

  15. Waste disposal from the light water reactor fuel cycle

    International Nuclear Information System (INIS)

    Costello, J.M.; Hardy, C.J.

    1981-05-01

    Alternative nuclear fuel cycles for support of light water reactors are described and wastes containing naturally occurring or artificially produced radioactivity reviewed. General principles and objectives in radioactive waste management are outlined, and methods for their practical application to fuel cycle wastes discussed. The paper concentrates upon management of wastes from upgrading processes of uranium hexafluoride manufacture and uranium enrichment, and, to a lesser extent, nuclear power reactor wastes. Some estimates of radiological dose commitments and health effects from nuclear power and fuel cycle wastes have been made for US conditions. These indicate that the major part of the radiological dose arises from uranium mining and milling, operation of nuclear reactors, and spent fuel reprocessing. However, the total dose from the fuel cycle is estimated to be only a small fraction of that from natural background radiation

  16. 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.)

  17. A model for oxidizing species concentrations in boiling water reactors

    International Nuclear Information System (INIS)

    Sun, B.; Chexal, B.; Pathania, R.; Chun, J.; Ballinger, R.; Abdollahian, D.

    1993-01-01

    To evaluate and control the intergranular stress corrosion cracking of boiling water reactor (BWR) vessel internal components requires knowledge of the concentration of oxidizing species that affects the electrochemical potentials in various regions of a BWR. In a BWR flow circuit, as water flows through the radiation field, the radiolysis process and chemical reactions lead to the production of species such as oxygen, hydrogen, and hydrogen peroxide. Since chemistry measurements are difficult inside BWRs, analytical tools have been developed by Ruiz and Lin, Ibe and Uchida and Chun and Ballinger for estimating the concentration of species that provide the necessary input for water chemistry control and material protection

  18. Nuclear calculation methods for light water moderated reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1961-02-01

    This report is intended as an introductory review. After a brief discussion of problems encountered in the nuclear design of water moderated reactors a comprehensive scheme of calculations is described. This scheme is based largely on theoretical methods and computer codes developed in the U.S.A. but some previously unreported developments made in this country are also described. It is shown that the effective reproduction factor of simple water moderated lattices may be estimated to an accuracy of approximately 1%. Methods for treating water gap flux peaking and control absorbers are presented in some detail, together with a brief discussion of temperature coefficients, void coefficients and burn-up problems. (author)

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

    International Nuclear Information System (INIS)

    Baindur, S.

    2008-01-01

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

  20. 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)

  1. Spiral-shaped reactor for water disinfection

    KAUST Repository

    Soukane, Sofiane

    2016-04-20

    Chlorine-based processes are still widely used for water disinfection. The disinfection process for municipal water consumption is usually carried out in large tanks, specifically designed to verify several hydraulic and disinfection criteria. The hydrodynamic behavior of contact tanks of different shapes, each with an approximate total volume of 50,000 m3, was analyzed by solving turbulent momentum transport equations with a computational fluid dynamics code, namely ANSYS fluent. Numerical experiments of a tracer pulse were performed for each design to generate flow through curves and investigate species residence time distribution for different inlet flow rates, ranging from 3 to 12 m3 s−1. A new nature-inspired Conch tank design whose shape follows an Archimedean spiral was then developed. The spiral design is shown to strongly outperform the other tanks’ designs for all the selected plug flow criteria with an enhancement in efficiency, less short circuiting, and an order of magnitude improvement in mixing and dispersion. Moreover, following the intensification philosophy, after 50% reduction in its size, the new design retains its properties and still gives far better results than the classical shapes.

  2. Fuel enrichment reduction for heavy water moderated research reactors

    International Nuclear Information System (INIS)

    McCulloch, D.B.

    1984-01-01

    Twelve heavy-water-moderated research reactors of significant power level (5 MW to 125 MW) currently operate in a number of countries, and use highly enriched uranium (HEU) fuel. Most of these reactors could in principle be converted to use uranium of lower enrichment, subject in some cases to the successful development and demonstration of new fuel materials and/or fuel element designs. It is, however, generally accepted as desirable that existing fuel element geometry be retained unaltered to minimise the capital costs and licensing difficulties associated with enrichment conversion. The high flux Australian reactor, HIFAR, at Lucas Heights, Sydney is one of 5 Dido-class reactors in the above group. It operates at 10 MW using 80% 235 U HEU fuel. Theoretical studies of neutronic, thermohydraulic and operational aspects of converting HIFAR to use fuels of reduced enrichment have been made over a period. It is concluded that with no change of fuel element geometry and no penalty in the present HEU fuel cycle burn-up performance, conversion to MEU (nominally 45% 235 U) would be feasible within the limits of current fully qualified U-Al fuel materials technology. There would be no significant, adverse effects on safety-related parameters (e.g. reactivity coefficients) and only small penalties in reactor flux. Conversion to LEU (nominally 20% 235 U) a similar basis would require that fuel materials of about 2.3 g U cm -3 be fully qualified, and would depress the in-core thermal neutron flux by about 15 per cent relative to HEU fuelling. In qualitative terms, similar conclusions would be expected to hold for a majority of the above heavy water moderated reactors. (author)

  3. Overview of activities for the reduction of dose rates in Swiss boiling water reactors

    International Nuclear Information System (INIS)

    Alder, H.P.; Schenker, E.

    1993-01-01

    Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

  4. Seawater desalination using small and medium light water reactors

    International Nuclear Information System (INIS)

    Shimamura, Kazuo

    2000-01-01

    Water is an essential substance for sustaining human life. As Japan is an island country, surrounded by the sea and having abundant rainfall, there is no scarcity of water in daily life except during abnormally dry summers or after disasters such as earthquakes. Consequently, there is hardly any demand for seawater desalination plants except on remote islands, Okinawa and a part of Kyushu. However, the IAEA has forecast a scarcity of drinking water in developing countries at the beginning of the 21st century. Further, much more irrigation water will be required every year to prevent cultivated areas from being lost by desertification. If developing countries were to produce such water by seawater desalination using current fossil fuel energy technology, it would cause increased air pollution and global warming. This paper explains the concept of seawater desalination plants using small and medium water reactors (hereinafter called 'nuclear desalination'), as well as important matters regarding the export nuclear desalination plants to developing countries. (author)

  5. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

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

  6. Long term review of research on light water reactor types

    International Nuclear Information System (INIS)

    Sumiya, Yutaka

    1982-01-01

    In Japan, 24 nuclear power plants of 17.18 million kWe capacity are in operation, and their rate of operation has shown the good result of more than 60% since 1980. One of the research on the development of light water reactors is the electric power common research, which was started in 1976, and 272 researches were carried out till 1982. It contributed to the counter-measures to stress corrosion cracking, thermal fatigue and the thinning of steam generator tubes, to the reduction of crud generation and the remote control and automation of inspection and maintenance, and to the verification of safety. The important items for the future are the cost down of nuclear power plant construction, the development of robots for nuclear power plants, the improvement of the ability to follow load variation, and the development of light water reactors of new types. It is necessary to diversify the types of reactors to avoid the effect of a serious trouble which may occur in one type of reactors. Tokyo Electric Power Co., Inc., thinks that the Japanese type PWRs having the technical features of KWU type PWRs are desirable for the future development. The compatibility with the condition of installation permission in Japan, the required design change and the economy of the standard design PWRs of KWU (1.3 million kW) have been studied since October, 1981, by KWU and three Japanese manufacturers. (Kako, I.)

  7. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  8. Qualification issues for advanced light-water reactor protection systems

    International Nuclear Information System (INIS)

    Korsah, K.; Clark, R.L.; Antonescu, C.

    1993-01-01

    The instrumentation and control (I ampersand C) systems in advanced reactors will make extensive use of digital controls, microprocessors, multiplexing, and fiber optic transmission. Elements of these advances in I ampersand C have been implemented on some current operating plants. However, the widespread use of the above technologies, as well as the use of artificial intelligence with minimum reliance on human operator control of reactors, highlights the need to develop standards for qualifying the I ampersand C used in the next generation of nuclear power plants. As a first step in this direction, the protection system I ampersand C for present-day plants was compared to that proposed for advanced light-water reactors (ALWRs). An evaluation template was developed by assembling a configuration of a safety channel instrument string for a generic ALWR, then comparing the impact of environmental stressors on that string to their effect on an equivalent instrument string from an existing light-water reactor. The template was then used to suggest a methodology for the qualification of microprocessor-based protection systems. The methodology identifies standards/regulatory guides (or lack thereof) for the qualification of microprocessor-based safety I ampersand C systems. This approach addresses in part issues raised in NRC policy document SECY-91-292, which recognizes that advanced I ampersand C systems for the nuclear industry are ''being developed without consensus standards. as the technology available for design is ahead of the technology that is well understood through experience and supported by application standards.''

  9. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Bouveret, F.

    2001-01-01

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  10. 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.

  11. Dense Medium Plasma Water Purification Reactor (DMP WaPR), Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — The Dense Medium Plasma Water Purification Reactor offers significant improvements over existing water purification technologies used in Advanced Life Support...

  12. Micropollutant removal from black water and grey water sludge in a UASB-GAC reactor.

    Science.gov (United States)

    Butkovskyi, A; Sevenou, L; Meulepas, R J W; Hernandez Leal, L; Zeeman, G; Rijnaarts, H H M

    2018-02-01

    The effect of granular activated carbon (GAC) addition on the removal of diclofenac, ibuprofen, metoprolol, galaxolide and triclosan in a up-flow anaerobic sludge blanket (UASB) reactor was studied. Prior to the reactor studies, batch experiments indicated that addition of activated carbon to UASB sludge can decrease micropollutant concentrations in both liquid phase and sludge. In continuous experiments, two UASB reactors were operated for 260 days at an HRT of 20 days, using a mixture of source separated black water and sludge from aerobic grey water treatment as influent. GAC (5.7 g per liter of reactor volume) was added to one of the reactors on day 138. No significant difference in COD removal and biogas production between reactors with and without GAC addition was observed. In the presence of GAC, fewer micropollutants were washed out with the effluent and a lower accumulation of micropollutants in sludge and particulate organic matter occurred, which is an advantage in micropollutant emission reduction from wastewater. However, the removal of micropollutants by adding GAC to a UASB reactor would require more activated carbon compared to effluent post-treatment. Additional research is needed to estimate the effect of bioregeneration on the lifetime of activated carbon in a UASB-GAC reactor.

  13. Identification of leaky steam generators by iodine mapping technique and development of tools for cutting of tubes of steam generators of Indian PHWRS

    International Nuclear Information System (INIS)

    Subba Rao, D.

    2006-01-01

    Kakrapar Atomic Power Station (2X220 MWe) located in Mandvi Taluka of Surat District in the state of Gujarat is the fifth Nuclear Power Station of the country. It has got an excellent record in the field of operation, safety, public awareness and emergency preparedness. KAPS Unit -1 achieved first criticality in Sep-1992 and was declared for commercial operation in may-1993. KAPS Unit -2 achieved first criticality in Jan-1995 and was declared for commercial operation in Sep-1995. So far station has generated about 30 billion units.Unit-1 achieved 98.4% and was graded as a world's No.1 in year 2002 amongst all CANDU type reactors. KAPS Unit -1 has made another record of operating continuously for more than 300 days in Indian PHWR s operating history. This paper mainly deals with the Indian PHWRs Steam Generators (SG) tube leaks, leaky steam generator identification by Iodine mapping, and development of special tool for cutting, removal and plugging of leaky tubes. These Steam Generators are designed by M/s Kraft Werke Union (KWU) of Siemens Group, West Germany, and Manufactured by M/s ENSA, SPAIN for Unit- 1 and by M/s MAN-GHH, Germany for Unit- 2. First time in October-2002 one of the Steam Generators of Unit-1 developed tube leak. To identify leaky Steam Generator, KAPS has established a method of Iodine mapping. With that the leaky SG was identified in very short time and corrective actions were taken immediately. Total three tube leaks (two in SG-4 of Unit-1 and one in SG-1 of unit-2) were experienced in both Units'. Following observations were made on SG tubes failure: All failures were in cold leg side; All Failures / deterioration locations were in front of main feed water nozzle; All Failures / deterioration locations were observed to be just above tube support plate (TSP) number 4 or 5; Deterioration ( i.e. wall thinning) observed from OD side and these tubes were adjacent to failed tubes; In all the three incidents, failed / deteriorated tubes were

  14. [Research of input water ratio's impact on the quality of effluent water from hydrolysis reactor].

    Science.gov (United States)

    Liang, Kang-Qiang; Xiong, Ya; Qi, Mao-Rong; Lin, Xiu-Jun; Zhu, Min; Song, Ying-Hao

    2012-11-01

    Based on high SS/BOD and low C/N ratio of waste water of municipal wastewater treatment plant, the structure of currently existing hydrolysis reactor was reformed to improve the influent quality. In order to strengthen the sludge hydrolysis and improve effluent water quality, two layers water distributors were set up so that the sludge hydrolysis zone was formed between the two layers distribution. For the purpose of the hydrolysis reactor not only plays the role of the primary sedimentation tank but also improves the effluent water biodegradability, input water ratios of the upper and lower water distributor in the experiment were changed to get the best input water ratio to guide the large-scale application of this sort hydrolysis reactor. Results show, four kinds of input water ratio have varying degrees COD and SS removal efficiency, however, input water ratio for 1 : 1 can substantially increase SCOD/COD ratio and VFA concentration of effluent water compared with the other three input water ratios. To improve the effluent biodegradability, input water ratio for 1 : 1 was chosen for the best input water ratio. That was the ratio of flow of upper distributor was 50%, and the ratio of the lower one was 50%, at this case it can reduce the processing burden of COD and SS for follow-up treatment, but also improve the biodegradability of the effluent.

  15. Unification of reactor elastomeric sealing based on material

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2012-01-01

    The unification of elastomeric sealing applications of Indian nuclear reactors based on a few qualified fluoroelastomer/perfluoroelastomer compounds and standardized approaches for finite element analysis (FEA) based design, manufacturing process and antifriction coatings is discussed. It is shown that the advance polymer architecture based Viton ® formulation developed for inflatable seals of 500 MWe Prototype Fast Breeder Reactor (PFBR) and its four basic variations can encompass other sealing applications of PFBR with minimum additional efforts on development and validation. Changing the blend ratio of Viton ® GBL 200S and 600S in inflatable seal formulation could extend its use to Pressurized Heavy Water Reactors (PHWRs). The higher operating temperature of Advanced Heavy Water Reactor (AHWR) seals expands the choice to perfluoroelastomers. FEA based on plane-strain/axisymmetric modeling (with Mooney–Rivlin as the basic constitutive model), seal manufacture by cold feed extrusion and injection molding as well as plasma Teflon-like coating belonging to two variations obtained from the development of inflatable seals provide the necessary standardization for unification. The gains in simplification of design, development and operation of seals along with the enhancements of safety and reliability are expected to be substantial.

  16. Economic competitiveness requirements for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  18. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  19. US Advanced Light Water Reactor Program; overall objective

    International Nuclear Information System (INIS)

    Klug, N.

    1989-01-01

    The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

  20. Fuel performance at high burnup for water reactors

    International Nuclear Information System (INIS)

    1991-02-01

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology. The purpose of this meeting was to review the ''state-of-the-art'' in the area of Fuel Performance at High Burnup for Water Reactors. Previous IAEA meetings on this topic were held in Mol in 1981 and 1984 and on related topics in Stockholm and Lyon in 1987. Fifty-five participants from 16 countries and two international organizations attended the meeting and 28 papers were presented and discussed. The papers were presented in five sub-sessions and during the meeting, working groups composed of the session chairmen and paper authors prepared the summary of each session with conclusions and recommendations for future work. A separate abstract was prepared for each of these papers. Refs, figs and tabs