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Sample records for mwe indian phwrs

  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. 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

  3. Evaluation of ultimate load bearing capacity of the primary containment of a typical first generation 220 MWe Indian PHWRs

    International Nuclear Information System (INIS)

    Singh, A.K.; Ray, I.; Roy, R.; Garg, R.P.; Verma, U.S.P.

    2005-01-01

    This paper presents the analysis of the Inner Containment Structure (ICS) of a typical first generation 220 MWe Indian PHWR for the purpose of evaluating its ultimate load bearing capacity (ULBC) beyond postulated design basis accident (DBA) scenario. The first generation ICS of Indian PHWRs are made of cylindrical wall capped with prestressed/reinforced cellular containment slab, which is connected monolithically to outer containment (OC) Wall to provide clamping effect during postulated DBA scenario. This paper discusses the simulation of construction sequence in analytical model, which is a very important aspect from point of view of capturing the residual stresses generated in the structure. The methodology adopted for the non-linear analysis of the prestressed concrete ICS including the various issues, viz. behaviour of concrete under compression and tension, tension stiffening, cracked shear modulus etc. have also been discussed in this paper. The effect of accident temperature on ULBC has been studied and discussed in this paper. This paper also discusses the study of mesh sensitivity of the finite element (FE) discretisation on ULBC of ICS in the non-linear range. Based on the detailed analysis, the factor of safety of the ICS beyond postulated DBA scenario has been evaluated. (authors)

  4. 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.

  5. Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe Indian PHWR: design perspective

    International Nuclear Information System (INIS)

    Sonavani, Manojkumar; Kelkar, M.G.; Singhvi, P.K.; Roy, S.; Ingle, V.J.

    2013-01-01

    The Flux Mapping System (FMS) of 700 MWe PHWR computes a detailed flux/power distribution of the reactor core using modal synthesis method and is also generate setback on different parameters by monitoring thermal neutron flux at more than 100 points inside the reactor core. These types of setbacks are introduced first time in Indian PHWRs. The paper brings out the Evolution of Flux Mapping System (FMS) from 540 MWe to 700 MWe and the overall design philosophy. The paper emphasizes on comparisons between 540 MWe and 700 MWe design, considerations for architectural design and setbacks for 700 MWe. (author)

  6. 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)

  7. 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)

  8. 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

  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. 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)

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. Design, development and deployment of special sealing plug for 540 MWe PHWRs

    International Nuclear Information System (INIS)

    Sharma, G.; Roy, S.; Patel, R.J.

    2012-01-01

    The coolant channel in Pressurized Heavy Water Reactors is a pressure boundary component and is very important for reactor performance and reactor safety. Monitoring the condition of the pressure tube of each coolant channel on a periodic basis is very important. In-Service Inspection (ISI) of the coolant channels in water filled condition is done regularly for 220 MWe PHWR. For the same purpose BARC Channel Inspection System is developed for 540 MWe PHWR also. Special Sealing Plug has been developed to facilitate the channel inspection (in water filled condition) with all necessary safety features at par with normal sealing plug. Special Sealing Plug provides a 50 mm through hole for passage of drive tube of Inspection Head maintaining integrity of PHT. Lot of challenges were faced for developing the Special Sealing Plug and its associated tools. It was a first of its kind design. First ISI of TAPS-4 was conducted successfully using this plug along with associated tools in November 2011. This development has provided immense help to NPCIL in life management of 540 MWe PHWR coolant channels. (author)

  17. PC based manual and safety logic card test setup for 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Chandgadkar, G.M.; Kohli, A.K.; Agarwal, R.G.; Chandra, Rajesh

    1992-01-01

    Fuel handling controls for 235 MWe PHWR make use of Manual and Logic cards (MLCs) for providing safety interlocks. These cards consist of various type of logic blocks. By connecting these logic blocks all the safety interlocks required for fuel handling controls have been provided. Previously trouble shooting of these cards was done by means of logic probe. Since the method was manual, it was laborious and time consuming. PC based test setup has overcome this drawback and detects the fault at the component level within few seconds. It also gives printout of status of faulty MLC cards. Here motherboard has been designed having slots for insertion of MLC cards. The input/output connection of these cards are coming to two 50 pin FRC connectors. PC communicates through 144 line digital input/output card with MLC card under test. Software is user friendly and outputs suitable input patterns to the card under test and checks for output pattern. It compares this output pattern with compare pattern and detects the fault and displays the symptoms. This system is currently in use at test facility for fuelling machine for 235 MWe PHWR reactor at Refuelling Technology Division, Hall-7. This test setup has been proposed for use at NAPP and future reactors. (author). 4 figs., 1 annexure

  18. 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)

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. 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

  4. 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

  5. Accident management-defence in depth in Indian PHWRS

    International Nuclear Information System (INIS)

    Jagannad, V.B.L.; Reddy, V.V.; Hajela, Sameer; Bhatia, C.M.; Nair, Suma

    2015-01-01

    Defence in Depth (DiD) is the established safety principle for the design of Nuclear Power Plants (NPPs). Accident at Fukushima Dai-ichi had highlighted the importance of provisions at Level-4 and 5 of DiD. Post Fukushima accident, on-site measures have been strengthened for Indian Nuclear Power Plants. On procedural front, Accident Management Guidelines have been introduced to handle events more severe than design basis accidents. This paper elaborates enhancement of Defence in Depth provisions for Indian Nuclear Power Plants. (author)

  6. 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

  7. Radiological safety related provisions and instrumentation in Indian PHWRs

    International Nuclear Information System (INIS)

    Ramamirtham, B.; Dabhadkar, S.B.; Sah, B.M.L.

    1994-01-01

    The collective radiation doses at the nuclear power plants (NPPs) world-wide have shown a significant downward trend which has resulted due to on-going efforts to keep exposures ALARA and also to meet the recently revised individual exposure limits of ICRP. In keeping with this trend a number of additional designed dose reduction features are also being incorporated in the Indian NPPs. These include better separation and shielding of radioactive systems/equipment, elimination of the use of cobalt-free materials in active systems, improved leak tightness of systems carrying heavy water, augmented ventilation and atmosphere drying systems, etc. The build-up of radiation levels in primary heat transport (PHT) system is controlled by incorporating improvements in the fuel performance and periodic system decontamination. Plant layouts have been modified, to improve the contamination control arrangements and optimum utilisation of dosimetry devices. For better control of internal exposures continuous efforts are on to make the protective gear more user-friendly. Green belts are being established around the NPPs to provide further protection against environmental impact. A number of additional radiation monitoring instruments /systems have been incorporated to provide information on radiation/activity levels, both within the plant and outside areas, particularly during emergency conditions. For processing of data provided by the large numbers of installed radiation instruments and initiating corrective/alarm actions, a computerised system (RADAS) has been provided. (author). 7 refs., 2 tabs

  8. 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)

  9. Power distribution monitoring and control in 500 MWe PHWR

    International Nuclear Information System (INIS)

    Kumar, A.

    1996-01-01

    The 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) is expected to be commissioned in a few years. It has a relatively large sized core with complex material distribution in comparison to the currently operating 220 MWe PHWRs. The resulting neutronically loosely coupled system demands continuous control of the core power distribution. This paper gives a brief description and analysis of the reactor monitoring and control system proposed for this reactor. (author). 11 refs, 8 figs, 3 tabs

  10. 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)

  11. 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.)

  12. 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)

  13. 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

  14. Analytical tools and methodologies for evaluation of residual life of contacting pressure tubes in the early generation of Indian PHWRs

    International Nuclear Information System (INIS)

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

    2002-01-01

    In-service life of a contacting Zircaloy-2 pressure tube (PT) in the earlier generation of Indian PHWRs, is limited mainly due to the accelerated hydrogen pick-up and nucleation and growth of hydride blister(s) at the cold spot(s) formed on outside surface of pressure tube as a result of its contact with the calandria tube (CT). The activities involving development of the analytical models for simulating the degradation mechanisms leading to PT-CT contact and the methodologies for the revaluation of their safe life under such condition form the important part of our extensive programme for the life management of contacting pressure tubes. Since after the PT-CT contact, rate of hydrogen pick-up and nucleation and growth of hydride blisters govern the safe residual life of the pressure tube, two analytical models (a) hydrogen pick-up model ('HYCON') and (b) model for the nucleation and growth of hydride blister at the contact spot ('BLIST -2D') have been developed in-house to estimate the extent of degradation caused by them. Along with them, a methodology for evaluation of safe residual life has also been formulated for evaluating the safe residual life of the contacting channels. This paper gives the brief description of the models and the methodologies relevant for the contacting Zircaloy-2 pressure tubes. (author)

  15. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    International Nuclear Information System (INIS)

    Sanatkumar, A.; Jit, I.; Muralidhar, G.

    1996-01-01

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs

  16. Fuel handling system of Indian 500 MWe PHWR-evolution and innovations

    Energy Technology Data Exchange (ETDEWEB)

    Sanatkumar, A; Jit, I; Muralidhar, G [Nuclear Power Corporation of India Ltd., Mumbai (India)

    1997-12-31

    India has gained rich experience in design, manufacture, testing, operation and maintenance of the Fuel Handling System of CANDU type PHWRs. When design and layout of the first 500 MWe PHWR was being evolved, it was possible for us to introduce many special and innovative features in the Fuel Handling System which are friendly for operations and maintenance personnel. Some of these are: Simple, robust and modular mechanisms for ease of maintenance; Shorter turnaround time for refuelling a channel by introduction of transit equipment between the Fuelling Machine (FM) Head and light water equipment; Optimised layout to transport spent fuel in straight and short path and also to facilitate direct wheeling out of the FM Head from the Reactor Building to the Service Building; Provision to operate the FM Head even when the Primary Heat Transport (PHT) System is open for maintenance; Control-console engineered for carrying out refuelling operations in the sitting position; and, Dedicated calibration and maintenance facility to facilitate quick replacement of the FM Head as a single unit. The above special features have been described in this paper. (author). 7 figs.

  17. 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

  18. Hydrogen related safety issues in the context of containments of Indian PHWRs

    International Nuclear Information System (INIS)

    Markendeya, S.G.; Ghosh, A.K.; Kushwaha, H.S.; Venkat Raj, V.

    2002-01-01

    Full text: Assessment of risk due to hydrogen released during postulated hypothetical accident scenarios in the nuclear power plants (NPPs) has been an important area of R and D studies world over. The issues, such as appropriate methodologies for estimation of hydrogen source term and for hydrogen dispersion calculations, technology development for hydrogen mitigation in containment of NPPs and assessment of damage due to deflagration/detonation of hydrogen (if it occurs) are being addressed as a part of some of the multidisciplinary study programs currently underway in BARC. While a significant overall progress has been achieved in general as a result of these programs, requirements of further fine-tuning of these studies have also emerged. The present paper takes a brief look at the current state-of the-art technology available to address these issues. The progress of R and D studies underway at BARC has also been critically reviewed in the paper to bring out necessary planning of further studies so as to enhance the safety of Indian NPPs

  19. Studies on flow induced vibration of reactivity devices of 700 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Prabhakaran, K.M., E-mail: kmprabha@yahoo.com [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Goyal, P.; Dutta, Anu; Bhasin, V.; Vaze, K.K.; Ghosh, A.K. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400 085 (India); Pillai, Ajith V.; Mathew, Jimmy [Nuclear Power Corporation of India Ltd., Mumbai 400 094 (India)

    2012-03-15

    Highlights: Black-Right-Pointing-Pointer FIV studies on internals of heavy water filled calandria of 700 MWe Indian PHWR is presented. Black-Right-Pointing-Pointer This includes CFD and structural dynamic analysis to predict the dynamic behavior of component lying inside calandria. Black-Right-Pointing-Pointer Results of these calculations as well as conclusions from this investigation are presented. Black-Right-Pointing-Pointer It is established that FIV is not a concern in the present design of calandria internals. - Abstract: Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. In the Indian nuclear industry, flow induced vibrations are assessed early in the design process and the results are incorporated in the design procedures. In this paper the details of flow induced vibration studies on internals like liquid zone control unit and poison injection units of heavy water filled calandria of 700 MWe Indian pressurized heavy water reactor is given. This includes computational fluid dynamics studies from which the velocities are extracted for the components lying inside the calandria. With these velocities as input, further studies are performed to predict the dynamic behavior of these components. Results of these calculations as well as conclusions derived from this investigation are presented. Based on the studies it has been established that flow induced vibration is not a concern in the present design of 700 MWe calandria internals.

  20. 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)

  1. 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)

  2. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

    2003-01-01

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  3. A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300{degrees}C. Two important observations of the experiments are - appreciable drop in maximum load at 300{degrees}C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis.

  4. A lead-before-break strategy for primary heat transport piping of 500 MWe Indian PHWR

    International Nuclear Information System (INIS)

    Chattopadhyay, J.; Dutta, B.K.; Kushwaha, H.S.

    1997-01-01

    Leak-Before-Break (LBB) is being used to design the primary heat transport piping system of 500 MWe Indian Pressurized Heavy Water Reactors (IPHWR). The work is categorized in three directions to demonstrate three levels of safety against sudden catastrophic break. Level 1 is inherent in the design procedure of piping system as per ASME Sec.III with a well defined factor of safety. Level 2 consists of fatigue crack growth study of a postulated part-through flaw at the inside surface of pipes. Level 3 is stability analysis of a postulated leakage size flaw under the maximum credible loading condition. Developmental work related to demonstration of level 2 and level 3 confidence is described in this paper. In a case study on fatigue crack growth on PHT straight pipes for level 2, negligible crack growth is predicted for the life of the reactor. For level 3 analysis, the R6 method has been adopted. A database to evaluate SIF of elbows with throughwall flaws under combined internal pressure and bending moment has been generated to provide one of the inputs for R6 method. The methodology of safety assessment of elbow using R6 method has been demonstrated for a typical pump discharge elbow. In this analysis, limit load of the cracked elbow has been determined by carrying out elasto-plastic finite element analysis. The limit load results compared well with those given by Miller. However, it requires further study to give a general form of limit load solution. On the experimental front, a set of small diameter pipe fracture experiments have been carried out at room temperature and 300 degrees C. Two important observations of the experiments are - appreciable drop in maximum load at 300 degrees C in case of SS pipes and out-of-plane crack growth in case of CS pipes. Experimental load deflection curves are finally compared with five J-estimation schemes predictions. A material database of PHT piping materials is also being generated for use in LBB analysis

  5. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-01-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  6. Evaluation of ultimate load bearing capacity of the primary containment of a typical 540 MWe Indian PHWR

    International Nuclear Information System (INIS)

    Ray, Indrajit; Roy, Raghupati; Verma, U.S.P.; Warudkar, A.S.

    2003-01-01

    This paper presents the analysis of the Inner Containment Structure (ICS) of a typical 540 MWe Indian PHWR for the purpose of evaluating its ultimate load bearing capacity (ULBC) under beyond postulated design basis accident (DBA) scenario. The methodology adopted for the non-linear analysis of the prestressed concrete ICS including the various issues, viz. behaviour of concrete under compression and tension, tension stiffening, cracked shear modulus etc. have also been discussed in this paper. The effect of accident temperature on ULBC has been studied and discussed in this paper. This paper also discusses about the study carried out for mesh sensitivity of the finite element (FE) discretization on ULBC of ICS in the non-linear range. Based on the detailed analysis, the factor of safety of the ICS under beyond postulated DBA scenario has been evaluated. (author)

  7. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Energy Technology Data Exchange (ETDEWEB)

    Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

    2005-08-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

  8. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh

    2005-01-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  9. 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)

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

    Directory of Open Access Journals (Sweden)

    A. Rama Rao

    2008-04-01

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

  11. Evolution in the design and development of the in-service inspection device for the Indian 500 MWe Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Singh, Ashutosh Pratap; Rajagopalan, C.; Rakesh, V.; Rajendran, S.; Venugopal, S.; Kasiviswanathan, K.V.; Jayakumar, T.

    2011-01-01

    Highlights: → Conceptual study on the configuration of an ISI device for FBR interspace environment has been carried out. → Prototyping of the concept has been experimentally validated in a mock up. → High temperature version of the ISI device has been made and tested in mock-up. Further experimentation is underway. → Simulation of different configurations of the device has been carried out with respect to reduced gap between main vessel and safety vessel for future FBRs. → Studies on wheel lining for the device have been carried out at 150 o C for better traction and payload capability. - Abstract: In-service inspection (ISI) plays a major role in monitoring the condition of nuclear power plant structures and components. Based on the information gathered during inspection and the studies carried out, it is possible to assess the extent of damage and take corrective measures to keep effects of ageing under control. In nuclear power plants comprehensive ISI is dictated by issues of increased safety to personnel and equipment, and efficiently enhances the plant life. A special emphasis has been laid on the development of robotic devices for the ISI of the indigenous Indian 500 MWe Prototype Fast Breeder Reactor (FBR) components. This paper traces the experiments and simulations in the key developments of a robotic device, for the ISI of main vessel and safety vessel of FBRs, carried out at Indira Gandhi Centre for Atomic Research, India.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  13. Supplementary shutdown system of 220 MWe standard PHWR in India

    International Nuclear Information System (INIS)

    Muktibodh, U.C.

    1997-01-01

    The design objective of the shutdown system is to make the reactor subcritical and hold it in that state for an extended period of time. This objective must be realised under all anticipated operational occurrences and postulated abnormal conditions even during most reactive state of the core. PHWR design criteria for shutdown stipulates requirement of two independent diverse and fast acting shutdown systems, either of which acting alone should meet the above objectives. This requirement would normally call for a large number of reactivity mechanism penetrations into the calandria. From the point of view of space availability at the reactivity mechanism area on top of calandria, for the relatively small core of 220 MWe PHWRs, and ease of maintenance realisation of the total worth by either of the shutdown systems acting alone was difficult. To overcome this engineering constraint and at the same time to satisfy the design criteria, a unique approach to meet the reactivity demands for shutdown was adopted. The reactivity requirements of the shutdown consists of fast and slow reactivity changes. For the shutdown system of 220 MWe PHWRs, the approach of realizing fast reactivity changes with dual redundant, diverse, fast acting shutdown systems aided by a slow acting shutdown system to counter delayed reactivity changes was conceived. The supplementary slow acting shutdown system is called upon to act after actuation of either of the two redundant fast acting systems and is referred to as Liquid Poison Injection System (LPIS). The system adds bulk amount of neutron poison (boric acid), equivalent to 45 mk, directly into the moderator through two nozzles in calandria using pneumatic pressure. This paper describes the design of LPIS as envisaged for the standardised 220 MWe PHWRs. (author)

  14. 3D core burnup studies in 500 MWe Indian prototype fast breeder reactor to attain enhanced core burnup

    International Nuclear Information System (INIS)

    Choudhry, Nakul; Riyas, A.; Devan, K.; Mohanakrishnan, P.

    2013-01-01

    Highlights: ► Enhanced burnup potential of existing prototype fast breeder reactor core is studied. ► By increasing the Pu enrichment, fuel burnup can be increased in existing PFBR core. ► Enhanced burnup increase economy and reduce load of fuel fabrication and reprocessing. ► Beginning of life reactivity is suppressed by increasing the number of diluents. ► Absorber rod worth requirements can be achieved by increasing 10 B enrichment. -- Abstract: Fast breeder reactors are capable of producing high fuel burnup because of higher internal breeding of fissile material and lesser parasitic capture of neutrons in the core. As these reactors need high fissile enrichment, high fuel burnup is desirable to be cost effective and to reduce the load on fuel reprocessing and fabrication plants. A pool type, liquid sodium cooled, mixed (Pu–U) oxide fueled 500 MWe prototype fast breeder reactor (PFBR), under construction at Kalpakkam is designed for a peak burnup of 100 GWd/t. This limitation on burnup is purely due to metallurgical properties of structural materials like clad and hexcan to withstand high neutron fluence, and not by the limitation on the excess reactivity available in the core. The 3D core burnup studies performed earlier for approach to equilibrium core of PFBR is continued to demonstrate the burnup potential of existing PFBR core. To increase the fuel burnup of PFBR, plutonium oxide enrichment is increased from 20.7%/27.7% to 22.1%/29.4% of core-1/core-2 which resulted in cycle length increase from 180 to 250 effective full power days (efpd), so that the peak fuel burnup increases from 100 to 134 GWd/t, keeping all the core parameters under allowed safety limits. Number of diluents subassemblies is increased from eight to twelve at beginning of life core to bring down the initial core excess reactivity. PFBR refueling is revised to accommodate twelve diluents. Increase of 10 B enrichment in control safety rods (CSRs) and diverse safety rods (DSRs

  15. Development of liquid poison injection system (SDS-2) for 500 MWe PHWRs

    International Nuclear Information System (INIS)

    Nawathe, Shirish; Umashankari, P.; Balakrishnan, Kamala; Mahajan, S.C.; Kakodkar, A.

    1991-01-01

    A secondary shut-down system (SDS-2) in the form of a mecahnism for introducing poison into the moderator of the PHWR is under development in Reactor Engineering Division of BARC. The system, as conceived, consists of a tank containing pressurised helium connected to poison tanks through quick opening solenoid valves. The tanks are connected to horizontal injection tubes in the calandria. On system actuation, gadolinium nitrate solution from the tanks passes to the injection tubes which have a number of holes through which the poison enters the moderator. This report details the development work being done on this poison injection system. An experimental facility was set up to measure the poison jet growth rate and the jet spread after injection, and mathematical models were developed to convert the observed jets into reactivity worth values. A description of the work and the computed results are presented. (author). 21 graphs. , 15 tabs

  16. Liquid poison injection system (LPIS) for KAIGA 1, 2 and RAPP 3, 4, 220 MWe PHWRs

    International Nuclear Information System (INIS)

    Soni, K.L.; Aparna; Mohan, L.R.; Nema, M.K.; Mahajan, S.C.

    1997-01-01

    LPIS is the modified version of existing Bulk Addition Mode (BAM) of the Automatic Liquid Poison Addition System (ALPAS). This BAM mode of ALPAS serves as slow acting supplementary shutdown system to PSS or SS as the case may be and hence it becomes a part of reactor safety system. The system (LPIS) has been envisaged to perform similar functions of ALPAS mode BAM. Now this is a self standing system. This feature of the system eliminates need of Gravity Addition of Boron System (GRAB). As the concept evolved is to be introduced for first time for the power reactors (from KAIGA-1, 2 and RAPP-3, 4 onwards), it has become necessary to check and verify its working by carrying out the necessary experimental and developmental work

  17. Development of channel inspection and gauging apparatus for 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Parulkar, S.K.; Taneja, R.; Taliyan, S.S.; Singh, Manjit; Govindarajan, G.

    1992-01-01

    Channel inspection and gauging apparatus is being developed to enable in-service channel inspection and gauging. Phase I apparatus to measure annular gap between pressure tube and calandria tube in a dry channel has been developed. The apparatus consists of a gauging head and a drive mechanism. The gauging head utilities an eddy current probe to measure the annular gap between pressure tube and calandria tube and an ultrasonic sensor to measure the wall thickness of the pressure tube. The output signal of the eddy current probe needs to be corrected for the effect of pressure tube wall thickness variation. This paper gives the details of the above apparatus. The results of calibration tests at mock-up station are presented. The paper outlines the program for the phase-wise development of Channel Inspection and Gauging Apparatus for use in heavy water filled channels without their isolation from PHT and draining. The final apparatus will have the facilities for ultrasonic flaw detection, ultrasonic gauging to measure pressure tube diameter and wall thickness, an inclinometer to measure slope and sag of pressure tube and eddy current probe for the measurement of annular gap between pressure tube and calandria tube. (author). 6 figs

  18. 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

  19. 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)

  20. Development of process route for production of tubing for various core sub-assemblies and heat exchangers for 500 MWe Indian PFBR

    International Nuclear Information System (INIS)

    Lakshminarayana, B.; Phani Babu, C.; Dubey, A.K.; Surender, A.; Deshpande, K.V.K.; Maity, P.K.

    2009-01-01

    India's three stage Nuclear Power Program has entered its second stage on commercial scale with the commencement of construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. Nuclear Fuel Complex (NFC), Hyderabad is playing a crucial role in the manufacture of all the critical sub-assemblies and control elements for this reactor. The challenging task of process development and production of the various critical tubing for these sub assemblies for PFBR has been taken up by Stainless Steel Tubes Plant (SSTP), NFC with indigenous development of the equipment and technology

  1. 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

  2. Flux mapping algorithm (FMA) for 700 MWe PHWR

    International Nuclear Information System (INIS)

    Sonavani, Manoj; Ingle, V.J.; Singhvi, P.K.; Raj, Manish; Fernando, M.P.S.; Kumar, A.N.

    2012-01-01

    For large reactor like 700 MWe PHWR effective spatial control is essential and is provided by RRS. For spatial control purpose reactor core is divided into 14 power zones. Corresponding to each zone is a light water zonal compartment. The 14 ZCCs are located in two radial planes, each containing 7 ZCCs. For each zone, power measurement is carried out using inconel (3 pitch long) self powered neutron detector (SPND) at appropriate location close to the respective ZCC. Since the zone power as obtained by the healthy zone control detector (ZCD) reading belonging to a particular zone may not correspond to its actual power because the detector per zone, measure only average fluxes but the zone extends over a large core region. Therefore accurate estimation of zone power calibration factors is required to estimate the zone powers and also to provide effective spatial power control to avoid the xenon induced spatial power oscillations in large PHWRs like 700 and 540 MWe Reactors. This accurate calculation of zone power is carried out by FMS which uses λ modes in its algorithm. Flux at any point inside the reactor can be represented in terms of the linear combination of these modes. Coefficients used in the expansion are called combining coefficient. If the readings of the detectors are known, then combining coefficients can be estimated by simple matrix operations. Once these combining coefficients are known, flux at any point inside the reactor can be found. (author)

  3. 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

  4. 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.

  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. Process Control Logic Modification to Mitigate Transient Following Tripping of a Primary Circulating Pump for a 540 MWe PHWR Power Plant

    International Nuclear Information System (INIS)

    Contractor, Ankur D; Gaikwad, Avinash J.; Kumar, Rajesh; Chakraborty, G.; Vhora, S.F.

    2006-01-01

    The 540 MWe Indian Pressurised Heavy Water Reactor (PHWR) incorporates many new features as compared to the earlier 220 MWe PHWRs. To evaluate the new design features like Primary Heat Transport (PHT) system configuration with two loops, four Primary Circulating Pumps (PCPs) and four passes through core, addition of a Pressurizer (surge Tank) in the PHT system along with Feed/Bleed system and their safety related implications, simulation model have been developed. A reactor step-back is proposed following one PCP trip. The corresponding PCP in the healthy loop is tripped to avoid asymmetrical flow and pressure distribution in the two identical loops. In spite of such elaborate provisions, the margins from high/low PHT pressure are small following tripping of one PCP. Mathematical models for all the major components and sub-systems of the proposed 540 MWe PHWR were developed based on the conservation equations of mass, momentum, energy and equation of state. All the associated control systems are also modeled. The PHT system includes the reactor core with nuclear fuel, PCP, PHT system pressure controller with feed/bleed system and Pressurizer (Surge Tank). The secondary system includes mainly the Steam Generators (SGs), the SG level and pressure controllers, apart from the various steam cycle components. All these models are integrated together to form the Plant Transient Analysis Computer Code Dyna540. The scenario following one PCP trips leads to different states (high/low pressure in Reactor Outlet Header (ROH)) depending upon the banks in which the PCP trips. The pressurizer is connected to two ROHs on one side of the reactor. The system pressure is controlled based on average of four ROHs pressure. In the case of asymmetrical pump operation, this logic leads to a situation where individual ROH pressure goes very near the low/high PHT system pressure trip set point, even though the controlled average pressure is very close to the set pressure. The PHT high

  7. 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)

  8. 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)

  9. 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)

  10. 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

  11. Use of gadolinium as neutron poison in 540 MWe PHWR

    International Nuclear Information System (INIS)

    Nag, P.K.; Fernando, M.P.S.; Kumar, A.N.

    2006-01-01

    In Pressurised heavy water reactors (PHWRs), neutron poison in the moderator is used to compensate the excess reactivity present in the core on different occasions such as xenon decay during synchronization just after poison out period or start ups from xenon free conditions. It is also used in secondary shutdown system (SDS-2), where required amount of neutron poison is injected directly into the moderator within 2.5 seconds. Further, it is also used for over poisoning the moderator to achieve the guaranteed shutdown state when the regular shutdown systems are taken for maintenance. Generally, two types of moderator poisons are used in power reactors to balance the reactivity of the core and they are boron and gadolinium. Gadolinium is used in the form of gadolinium nitrate (Gd(NO 3 ) 3 .6H 2 O). The paper gives the details of estimation of reactivity coefficients of gadolinium for 540 MWe PHWR for different operating conditions. These neutron poisons are converted into non-absorbing elements and therefore their effective worth will decrease as reactor operation proceeds. The rate of burning of neutron absorbing isotopes depends on its magnitude of absorption cross-section and thermal flux seen by them. The present study discusses the burning characteristics of gadolinium during power operation in 540 MWe PHWR. It is established by detailed analysis that the rate of positive reactivity realized due to burning of neutron absorbing Gd isotopes almost match with the build up rate of xenon. The burning half lives of boron and gadolinium is worked out for different power levels. (author)

  12. Three-dimensional studies of the 700 MWe steam generator design

    International Nuclear Information System (INIS)

    John, B.; Pietralik, J.

    2006-01-01

    The next stage in the Indian nuclear power programme envisions building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment including the reactor, steam generators (SGs), turbo-generator, major pumps, etc. The SG used in the current generation of 540 MWe IPHWRs, is a mushroom type, inverted U-tube, natural-circulation SG. The 700 MWe SG is of the same type and has the same tube bundle design and the same heat transfer area. The tube diameter, tube pitch, and outer diameter of the SG sections are the same as for the 540 MWe SG. The geometry of the feedwater header, the flow restrictor in the downcomer and the flow distribution plate are different in the two designs. The changes were required due to a 26% increase in steam flow rate while maintaining the same circulation ratio. This paper describes the design of the 700 MWe SG and a thermalhydraulic analysis using a one-dimensional, in-house code and a three-dimensional code called THIRST developed by AECL. The codes were validated against the 540 MWe SG data. The analysis was made for the 700 MWe SG for two versions: with and without integral preheater. The results of the THIRST runs were used for a flow-induced vibration analysis. The results of the flow-induced vibration analysis show that the vibrations are not excessive. (author)

  13. Development of innovative tools based on fuelling machine for ageing management of coolant channels of 220 MWe PHWRs

    International Nuclear Information System (INIS)

    Dev, Mahender; Roy, Shyamal; Bhattachrya, Sambit; Singh, Jit Pal; Patel, R.J.; Agarwal, R.G.

    2006-01-01

    PHWR coolant channels are required to be inspected periodically to satisfy the regulatory requirement, to provide information about known or suspected problem and to provide information to assist in future design. This paper describes these tools and techniques, their capabilities and experience of implementing these in reactor site

  14. 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

  15. Design of shutdown system no.2 liquid poison injection system for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Bhatnagar, S.; Balasubrahmanian, A.K.; Pillai, A.V.

    1997-01-01

    Defence in depth and two group system concepts form the basic design philosophy for the shutdown systems. There are two independent, diverse and fast acting shutdown systems provided for the 500 MWe PHWR. The design is based on fail-safe principle, sufficient component redundancy and on-line testing. Liquid poison injection system, as shutdown system 2, is newly developed for the 500 MWe PHWRs. The system operates by rapidly injecting gadolinium nitrate solution into bulk moderator using stored helium pressure thereby inserting negative reactivity. A high pressure helium supply tank which provides the energy for system actuation, is connected, through an array of fast acting valves in series-parallel arrangement, to the individual poison tanks storing gadolinium nitrate solution. The valves, belonging to three different channels of reactor Protection System 2, are the only active components in the system. The valves are fail safe and are periodically tested on-line without actually firing the system. The system comprising of in-core assemblies and the external process system has been engineered. Experimental work is being carried out by BARC for design validation and data generation. This paper describes the conceptual development, design basis, design parameters and detailed engineering of the system. (author)

  16. Improved 1500 MWe Arabelle begins operation

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    Two of the first 1500 MWe steam turbine-generator sets, described as the largest in the world, are undergoing commissioning at the Chooz B PWR nuclear power station in France. A number of design improvements have been made over the previous generation of 1350 MWe turbines, a process which will continue. (Author)

  17. A review of the Indian fast reactor programme

    International Nuclear Information System (INIS)

    Paranjpe, S.R.; Bhoje, S.B.

    1991-01-01

    Production of electricity during April 1990 - March 1991 was 200 TWh with an increase of 7% over last year. Contribution from coal based thermal is 70%, nuclear 2.5% and 27.5 from hydro. Electricity demand is increasing more than the production growth rate. The programme of installation of 10,000 MWe nuclear capacity in PHWRs by the year 2000 is in progress. 8 x 235 MWe PHWRs are under commissioning and construction. The Government has sanctioned construction of the first 2 x 500 MWe PHWR. Progress in the construction of NPPs is somewhat slow due to industrial infrastructure and financial constraints. There is no public opposition to nuclear power. An intergovernment agreement has been singed between India and the USSR for construction of 2 x 1000 MWe PWRs. FBTR is being operated intermittently up to a power level of 1 MWt without steam generators. Power operation is delayed due to commissioning of a hydrogen leak detection system for the steam generators. (author)

  18. Insights gained from PSAs of French 900MWe and 1300MWe units

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.-M.; Villemeur, A.; Berger, J.-P.; Guio, J.-M. de

    1991-01-01

    The two probabilistic safety assessments of 900MWe and 1300MWe Pressurized Water Reactors (PWRs) recently completed in France constitute an important knowledge resource for the assessment of PWR safety. One innovative feature of this research programme, which yielded many valuable lessons, comes from the fact that plant shutdown state and long term post-accident conditions were fully taken into account. (author)

  19. Monitoring-control of the 900 MWe and 1300 MWe nuclear reactors

    International Nuclear Information System (INIS)

    Meyer, J.

    1982-01-01

    After a short definition of the monitoring-control of the 900 MWe and 1300 MWe nuclear reactors, and a recall of requirements of nuclear energy, this paper presents the following points concerning the whole system of monitoring-control: the organization, the systems (instrumentation, automation), the technologies, the imperfections and the improvements brought to the system [fr

  20. 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

  1. CAREM project 15 to 150 MWe

    International Nuclear Information System (INIS)

    1992-05-01

    The main goal of the Carem Project is the introduction of a Inherent Safe Nuclear Power Reactor in the range of low power (15 to 150 MWe). For this low-power application, light-water and low enriched uranium was selected, since using those concepts permits to take full advantage of the special characteristics of low power reactors. INVAP has been involved in the last years in the design and construction of a Carem Reactor, which could cover a range up to 150 MWe, using a multiple-unit approach. It would furnished the 150 MWe, using six Carem Reactors, of proper power, which would share most of the services. INVAP is a reliable supplier of not only the nuclear reactor but also of the fuel

  2. Evolution of MMI for 500 MWe PHWR plant

    International Nuclear Information System (INIS)

    Surendar, Ch.; Sharma, M.P.; Jayanthi, S.

    1994-01-01

    The Indian nuclear power programme for building Pressurized Heavy Water Reactors began with the construction of two units at Kota, Rajasthan. Although the concept of a centralized control room has been used since the beginning, the man-machine interface design has evolved with technological developments. The man-machine interaction in the earliest plants imposed a considerable burden on the operators and led to a need for more sophisticated instrumentation. Several microprocessor and computer based systems were identified and developed and many were retrofitted into existing plants providing immediate advantages. This paper traces the evolution of many of these systems and also describes the basis and the architecture for the man-machine interaction scheme in the 500 MWe nuclear power plants currently being designed. (author). 7 refs., 2 figs., 1 tab

  3. French 900 MWe PWR PSA preliminary results

    International Nuclear Information System (INIS)

    Lanore, J.M.; Brisbois, J.

    1988-10-01

    A PSA is performed by the Safety Assessment Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the relative preliminary results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures

  4. 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

  5. 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)

  6. Seismic qualification of moderator system pump-motor units for RAPP-3,4 and KAIGA-1,2 235 MWe PHWRs

    International Nuclear Information System (INIS)

    Neelwarne, A.; Soni, R.S.; Kushawaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1992-01-01

    Smooth operation of active components like primary heat transport pumps, moderator pumps, emergency core cooling pumps etc. is always required to ensure safety of any nuclear power plants in case of normal as well as abnormal conditions such as earthquake loading. In order to ensure the functional requirement of such rotating equipment, is necessary to demonstrate, either through theoretical means or through experimental means, that in an event like earthquake loading, the static parts and the rotating parts of the equipment do not rub against each other giving rise to trouble during their operation. The moderator system pump units for RAPP-3,4 and Kaiga-1,2 have been analysed theoretically to demonstrate the structural integrity of various components of the unit as well as the functional requirement during an earthquake loading. A detailed Finite Element Model (FEM) was prepared for this which includes the modelling of static parts, rotating parts, anti-friction bearings and fluid-film journal bearings. Response spectrum analysis of the unit was carried out using the applicable floor response spectra for RAPP-3,4 and Kaiga-1,2 sites. It was concluded from this analysis that the pump-motor unit analysed meets the required design intent in terms of structural integrity and operability of the unit. The present report gives a detailed description of the problem, the development of FEM model, results and the conclusions arrived at. (author). 23 refs., 9 tabs., 17 figs

  7. Water treatment for 500 MWe PHWR plants

    International Nuclear Information System (INIS)

    Vasist, Sudheer; Sharma, M.C.; Agarwal, N.K.

    1995-01-01

    Large quantities of treated water is required for power generation. For a typical 500 MWe PHWR inland station with cooling towers, raw water at the rate of 6000 m 3 /hr is required. Impurities in cooling water give rise to the problems of corrosion, scaling, microbiological contamination, fouling, silical deposition etc. These problems lead to increased maintenance cost, reduced heat transfer efficiency, and possible production cut backs or shutdowns. The problems in coastal based power plants are more serious because of the highly corrosive nature of sea water used for cooling. An overview of the cooling water systems and water treatment method is enumerated. (author). 2 refs., 1 fig

  8. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

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

  9. 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)

  10. Safety parameter display system: an operator support system for enhancement of safety in Indian PHWRs

    International Nuclear Information System (INIS)

    Subramaniam, K.; Biswas, T.

    1994-01-01

    Ensuring operational safety in nuclear power plants is important as operator errors are observed to contribute significantly to the occurrence of accidents. Computerized operator support systems, which process and structure information, can help operators during both normal and transient conditions, and thereby enhance safety and aid effective response to emergency conditions. An important operator aid being developed and described in this paper, is the safety parameter display system (SPDS). The SPDS is an event-independent, symptom-based operator aid for safety monitoring. Knowledge-based systems can provide operators with an improved quality of information. An information processing model of a knowledge based operator support system (KBOSS) developed for emergency conditions using an expert system shell is also presented. The paper concludes with a discussion of the design issues involved in the use of a knowledge based systems for real time safety monitoring and fault diagnosis. (author). 8 refs., 4 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  12. Towards commercial fast breeder reactors the first 1200 MWe unit

    International Nuclear Information System (INIS)

    Banal, M.; Carle, R.

    The public probably thinks of these fast breeder reactors in terms of their rising unit capacity: RAPSODIE 20 MW (thermal), raised to 40 MW, PHENIX 25 MWe, and now 1200 MWe. However, the purposes of the project and the framework of construction have been fundamentally different in each case. Design parameters and the development program of the LMFBR are presented. (auth)

  13. Safety options for the 1300 MWe program

    International Nuclear Information System (INIS)

    Cayol, A.; Dupuis, M.C.; Fourest, B.; Oury, J.M.

    1980-04-01

    Standardization of the nuclear plants built in France implies an examination of the main technical safety options to be taken for a given type of reactor. By this procedure the subjects for which detailed studies will be needed to confirm the decisions made for the project can be defined in advance. In this context the technical safety option analysis for the 1300 MWe plants was conducted from the end of 1975 to the middle of 1978 according to usual regulation examination practice. The main conclusions are presented on the following subjects: safety methods; technical options concerning the containment vessel, primary fluid activity, fuel elements, steam generators; general organization of the lay-out [fr

  14. Safety margin improvement by adopting the feature of interleaving in 700 MWe PHWR

    International Nuclear Information System (INIS)

    Kumar, Nrependra; Yadav, S.K.; Khan, T.A.; Dixit, A.; Singhal, Mukesh; Nair, Suma R.

    2015-01-01

    Indian Pressurised Heavy Water Reactors (IPHWRs) of 700 MWe are under construction at Kakrapar Atomic Power Project -3,4 and Rajasthan Atomic Power Project-7,8. These units have enhanced safety features with respect to standard IPHWRs. One of the enhanced features is interleaving of feeders/channels. In interleaved feeder configuration, each header located at either end of reactor gets connected to one quarter of core channels, which are uniformly distributed. The core is divided into two loops with feeder connected in interleaved fashioned. In this paper a comparative study has been performed between the two cases: 1) The core splits in two vertical halves and each vertical half is a loop of PHT (TAPS-3 and 4 Type configuration). 2) The core is divided into two loops with feeders/ channels connected in interleaved fashioned (700 MWe Configuration). LOCA studies have been performed for 700 MWe PHWR considering interleaving of feeders configuration using in-house developed computer code ATMIKA and 3-D neutron kinetics code IQS-3D. The issue of interleaving is closely linked to an inherent reactivity characteristic of PHWR reactors (viz., positive void reactivity coefficient) which leads to a power increase following a Large LOCA. In 700 MWe PHWR with intent to improve the safety margin, adopted the feature of interleaving of feeders which causes in reduction in the magnitude of void coefficient and results in reduction of peak power during LBLOCA. The systematic LBLOCA study demonstrates that interleaved configuration of feeder/channels of two loops has higher safety margins (i.e. with respect to peak power, prompt-criticality margin, adiabatic heat deposition on the fuel pins, sheath temperature excursion and clad oxidation) with regard to the effectiveness of shutdown system. (author)

  15. Paluel: the first of the 1300 MWe class

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The 1300 MWe class follows that of 900 MWe class. It is the result of studies which have taken into account the evolution of projects made by manufacturers, of research into economies of scale and site optimisation and the attempt to secure a reputation in the export field. In comparison with the 900 MWe class, the 1300 MWe class offers both similarities and differences. Similarities: the general design of the pressure vessel and the fuel elements is the same, as is the design of the loops on the primary cooling circuit. With the aim of reducing costs, the Equipment department carried out a study in 1978 regarding a number of slight modifications in the design called P'4, consisting of at least 14 units, orders for which will be given in the period up to 1983-84 [fr

  16. The THESEUS project -- 50 MWe solar thermal power for Crete

    Energy Technology Data Exchange (ETDEWEB)

    Schillig, F.; Geyer, M.; Kistner, R.; Aringhoff, R.; Nava, P.; Brakmann, G.

    1998-07-01

    A consortium of European industry, utilities and research institutions from Greece, Germany, Spain and Italy attempts to implement a 52 MWe solar thermal power plant with parabolic trough technology on the Greek island of Crete sponsored by the EU' s THERMIE program. The increased demand for electricity on the island, a consequence of the growing allurement of the island as a tourist resort, makes it necessary to expand the installed capacity on Crete during the next years. According to the capacity expansion plans of Greek' s utility PPC a 160 MWe heavy fuel-fired power plant complex--two 30 MWe diesel units and two 50 MWe steam turbine units--is foreseen to be built by the year 2002. In this paper a description of the technical, economical and environmental aspects of the THESEUS project is provided. Moreover a market entry strategy for solar thermal power generation is discussed.

  17. Listing of nuclear power plant larger than 100 MWe

    International Nuclear Information System (INIS)

    McHugh, B.

    1976-03-01

    This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed by NSSS supply. (M.S.)

  18. 300 MWe Burner Core Design with two Enrichment Zoning

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Sang Ji; Kim, Yeong Il

    2008-01-01

    KAERI has been developing the KALIMER-600 core design with a breakeven fissile conversion ratio. The core is loaded with a ternary metallic fuel (TRU-U-10Zr), and the breakeven characteristics are achieved without any blanket assembly. As an alternative plan, a KALIMER-600 burner core design has been also performed. In the early stage of the development of a fast reactor, the main purpose is an economical use of a uranium resource but nowadays in addition to the maximum utilization of a uranium resource, the burning of a high level radioactive waste is taken as an additional interest for the harmony of the environment. In way of constructing the commercial size reactor which has the power level ranging from 800 MWe to 1600 MWe, the demonstration reactor which has the power level ranging from 200 MWe to 600 MWe was usually constructed for the midterm stage to commercial size reactor. In this paper, a 300 MWe burner core design was performed with purpose of demonstration reactor for KALIMER-600 burner of 600 MWe. As a means to flatten the power distribution, instead of a single fuel enrichment scheme adapted in design of KALIMER-600 burner, the 2 enrichment zoning approach was adapted

  19. Use of artificial neural network in estimating channel power distribution of a 220 MWe PHWR

    International Nuclear Information System (INIS)

    Dubey, B.P.; Chandra, A.K.; Govindarajan, G.; Jagannathan, V.; Kataria, S.K.

    1998-01-01

    Knowledge of the distribution of power in all the 306 channels of a Pressurised Heavy Water Reactor (PHWR) as a result of the movement of one or more of the four regulating rods is important from the operation and maintenance point view of the reactor. Conventional computer codes available for this purpose take several minutes to calculate the channel power distribution on PC-AT/486. An Artificial Neural network (ANN), based on the RPROP algorithm has been developed and employed in predicting channel power distribution of a 220 MWe Indian PHWR as a result of a perturbation caused by the movement of one or more of the four regulating rods of the reactor. The ANN based system produces the result of an analysis much faster than that produced by a conventional computer code usually employed for this application. The ANN based system is rugged, accurate and fast, and therefore, has potential to be used in real-time applications. (author)

  20. 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

  1. Instrumentation of steam cycle HTR's up to 900 MWe

    International Nuclear Information System (INIS)

    Leithner, D.E.; Winkenbach, B.

    1982-06-01

    Due to basic design features and inherent safety qualities in-core instrumentation is not needed in an HTR. Reactor safety requirements can be met by integral measurements. A modest spatial resolving power of the out-of-core instrumentation is sufficient for all operational purposes in small and medium sized steam cycle HTR's. Thus, the instrumentation concept of the THTR 300 MWe prototype reactor can be adopted without major changes for the HTR 450 MWe reactor project, as is demonstrated here for the neutron flux and temperature measurements. (author)

  2. RHTF 2, a 1200 MWe high temperature reactor

    International Nuclear Information System (INIS)

    Brisbois, Jacques

    1978-01-01

    After having adapted to French conditions the 1160 MWe G.A.C. reactor, Commissariat a l'Energie Atomique and French Industry have decided to design an High Temperature Reactor 1200 MWe based on the G.A.C. technology and taking into account the point of view of Electricite de France and the experience of C.E.A. and industry on the gas cooled reactor technology. The main objective of this work is to produce a reactor design having a low technical risk, good operability, with an emphasis on the safety aspects easing the licensing problems

  3. Experience on KKNPP VVER 1000 MWe water chemistry

    International Nuclear Information System (INIS)

    Ganesh, S.; Selvaraj, S.; Balasubramanian, M.R.; Selvavinayagam, P.; Pillai, Suresh Kumar

    2015-01-01

    Kudankulam Nuclear Power Project consists of pressurized water reactor (VVER) 2 x 1000 MWe constructed in collaboration with Russian Federation at Kudankulam in Tirunelveli District, Tamilnadu. Unit - 1 attained criticality on July 13 th 2013 and the unit was synchronized to grid on 22 nd October 2013. This paper highlights experience gained on water chemistry regime for primary and secondary circuit. (author)

  4. The french 900 MWe PWR PSA results and specificities

    International Nuclear Information System (INIS)

    Lanore, J.M.

    1990-01-01

    A probabilistic Safety Assessment has been performed by the Safety Analysis Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the level 1 results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures and the importance of risk during shutdown situations

  5. Listing of nuclear power plant larger than 100 MWe

    International Nuclear Information System (INIS)

    McHugh, B.

    1975-06-01

    This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed alphabetically. The report contains also a plant ranking list, where the plants are listed by the load factor (12 months) (M.S.)

  6. Listing of nuclear power plant larger than 100 MWe

    International Nuclear Information System (INIS)

    McHugh, B.

    1975-12-01

    This report contains a list of all nuclear power plants larger than 100 MWe, printed out from the Argus Data Bank at Chalmers University of Technology in Sweden. The plants are listed by country. The report contains also a plant ranking list, where the plants are listed by the load factor (12 months). (M.S.)

  7. Improved design on Qinshan 300 MWe nuclear power plant

    International Nuclear Information System (INIS)

    Shi Peihua; Cheng Wanli; Lu Rongliang

    1993-01-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented

  8. Improved design on Qinshan 300 MWe nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Peihua, Shi; Wanli, Cheng; Rongliang, Lu [Shanghai Nuclear Engineering Research and Design Inst. (China)

    1993-06-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented.

  9. Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

    International Nuclear Information System (INIS)

    Kapoor, K.

    2015-01-01

    Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

  10. A multinode digital control system for 500 MWe PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Patil, G N; Suresh Babu, R M; Jangra, L R; Das, Shantanu; Mallik, S B [Bhabha Atomic Research Centre, Bombay (India). Reactor Control Div.

    1994-12-31

    A fault tolerant distributed digital computer system for 500 MWe reactor power regulation is configured around standard microcomputer boards designed indigenously. The system is configured as functionally partitioned distributed control system having 8 nodes linked by high-speed dual redundant high-way. The paper gives the details of the configuration of system and how the features of fault-tolerance and fail-safeness are achieved through design. (author). 1 fig.

  11. A 1500-MW(e) HTGR nuclear generating station

    International Nuclear Information System (INIS)

    Stinson, R.C.; Hornbuckle, J.D.; Wilson, W.H.

    1976-01-01

    A conceptual design of a 1500-MW(e) HTGR nuclear generating station is described. The design concept was developed under a three-party arrangement among General Atomic Company as nuclear steam supply system (NSSS) supplier, Bechtel Power Corporation as engineer-constructors of the balance of plant (BOP), and Southern California Edison Company as a potential utility user. A typical site in the lower Mojave Desert in southeastern California was assumed for the purpose of establishing the basic site criteria. Various alternative steam cycles, prestressed concrete reactor vessel (PCRV) and component arrangements, fuel-handling concepts, and BOP layouts were developed and investigated in a programme designed to lead to an economic plant design. The paper describes the NSSS and BOP designs, the general plant arrangement and a description of the site and its unique characteristics. The elements of the design are: the use of four steam generators that are twice the capacity of GA's steam generators for its 770-MW(e) and 1100-MW(e) units; the rearrangement of steam and feedwater piping and support within the PCRV; the elimination of the PCRV star foundation to reduce the overall height of the containment building as well as of the PCRV; a revised fuel-handling concept which permits the use of a simplified, grade-level fuel storage pool; a plant arrangement that permits a substantial reduction in the penetration structure around the containment while still minimizing the lengths of cable and piping runs; and the use of two tandem-compound turbine generators. Plant design bases are discussed, and events leading to the changes in concept from the reference 8-loop PCRV 1500-MW(e) HTGR unit are described. (author)

  12. Forbush decreases observed 40 mwe underground in 1978

    International Nuclear Information System (INIS)

    Benko, G.; Kecskemety, K.; Neuprandt, G.; Somogyi, A.J.

    1982-01-01

    Forbush decreases observed 40 mwe underground at Budapest in the first half of 1978 have been analysed together with the data of several neutron monitor stations in Europe. Assuming a power-exponential type spectrum for the variations spectrum in space as a function of rigidity, the best fitting values of power and upper cut-off rigidity have been calculated from maximum decrement by means of the weighted least squares method

  13. Dissolved oxygen removal on radiolysis: studies in context of use of nitrogen atmosphere above PHT storage bag in Indian PHWRs

    International Nuclear Information System (INIS)

    Kumbhar, A.G.; Venkateswaran, G.; Kishore, K.; Kumar, Sangeeta D.; Naik, D.B.

    2008-01-01

    Dissolved oxygen content of the water (N 2 in gas phase) sample on radiolysis was measured and it was observed that up to 2 M Rad dose, oxygen content decreases linearly and at higher doses remains constant. Results are compared with nitrate ion yield in water-N 2 systems determined earlier. In aerated solutions also, nitrate ion yield was measured as function of dose. (author)

  14. Dynamic response of domes in CANDU 600 MWe containments

    International Nuclear Information System (INIS)

    Aziz, T.S.; Meng, V.; Alizadeh, A.

    1981-01-01

    CANDU reactors of the 600 MWe type are typically housed in a cylindrical prestressed concrete containment structure; rising from a flat slab and ending in a domed roof. The principal components of this structure are: (a) a circular base slab, (b) a vertical cylinder and (c) a spherical dome cap. A unique feature of a CANDU 600 MWe containment structure is the existence of an inner spherical concrete dome, located below the outer spherical dome, which serves as the bottom of a reservoir for the storage of 560,000 imperial gallons of douzing water. The thickness of the prestressed cylinder wall is approximately doubled between the two domes to create a ring beam. Inside the containment there exists an internal concrete structure which is independent of the containment structure except for support on the base slab. The containment boundary is a fully prestressed concrete structure. This paper deals with the seismic behaviour of the CANDU 600 MWe containment structure and the effect of its unique features; such as the lower dome and the douzing water on this behaviour. The objective of the study is to evaluate the interaction (coupling) effects between the different components of the structure. The approach taken is to study each component of the structure individually, then an assembly of the different components, and finally the total containment structure. This presentation is limited to the vertical response of the structure under a vertical earthquake only. Axisymmetric finite elements were used in all models. The vertical responses at selected points of the structure were obtained by the response spectrum method as well as the time-history method. It was observed that the response spectrum method over-estimates the vertical response of the domes and under-estimates the vertical responses of the ring girder and the containment cylinder compared to the time-history method. (orig./RW)

  15. A 600 MWe advanced PWR for the 1990's

    International Nuclear Information System (INIS)

    Lemon, J.E.; Malloy, J.D.; Allen, R.E.

    1987-01-01

    The Babcock and Wilcox Company (B and W) and United Engineers and Constructors (UE and C) have prepared a conceptual design of an advanced 600 MWe Presurized Water Nuclear Power Plant. This design utilizes the large body of design and operating experience on PWRs in the U.S. and abroad and incorporates improvements emphasizing simplicity, safety, licensability, ease of construction, operability, reliability and maintainability. Cost and schedule estimates based on U.S. utility experience indicate that this plant design should be competitive with alternate options

  16. The status of safeguarding 600 MW(e) CANDU reactors

    International Nuclear Information System (INIS)

    Von Baeckmann, A.; Rundquist, D.E.; Pushkarjov, V.; Smith, R.M.; Zarecki, C.W.

    1982-09-01

    There has been extensive work in the development of CANDU safeguards since the last International Conference on Nuclear Power, and this has resulted in the development of improved equipment for the safeguards system now being installed in the 600 MW(e) CANDU generating stations. The overall system is designed to improve on the existing IAEA safeguards and to provide adequate coverage for each plausible nuclear material diversion route. There is sufficient sensitivity and redundancy to enable the timely detection of the possible diversion of significant quantities of nuclear material

  17. Technical feasibility study of 60 MWe fast reactor concept: RAPID

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Ueda, Nobuyuki; Uotani, Masaki

    1993-01-01

    A study has been performed on the passive safety features and technical feasibility of an inherently safe 60 MWe fast reactor concept RAPID to meet various power requirements in Japan. The system dynamic analyses on the UTOP and ULOF transients revealed that the enhanced reactivity feedback derived from an annular core configuration and the integrated fuel assembly provides a high margin of self-protection. Structural integrity of the integrated fuel assembly has also been confirmed. The following innovative key technologies have been demonstrated; Lithium Injection Modules (LIM) for ultimate shutdown, Lithium Expansion Modulus (LEM) for inherent reactivity feedback and Void Leading Channel (VLC) for the sodium void worth reduction. (author)

  18. Potential use of thorium through fusion breeders in the Indian context

    International Nuclear Information System (INIS)

    Srinivasan, M.; Basu, T.K.; Subba Rao, K.

    1991-01-01

    The Indian Nuclear Programme is based on a three stage strategy: the first stage of about 10 GWe comprises of natural uranium fuelled Pressurised Heavy Water Reactors (PHWRs); the second stage would consist of Liquid Metal Cooled Fast Breeder Reactors (LMFBRs) to be fuelled with plutonium generated in the first stage PHWRs and the third stage is envisaged to be based on advanced converters/breeders operating on the Th/U-233 cycle. It has generally been assumed that the initial inventory of U-233 for the third stage reactors would be generated in the blankets of LMFBRs containing thorium. But the success of this strategy depends crucially on the attainment of LMFBR doubling times as short as 14 years. The progress registered in recent years in the magnetic confinement of fusion plasmas has opened up the prospects of developing Fusion Breeders for the direct conversion of fertile 232 Th into fissile 233 U using the 14 MeV neutron released in the (D-T) fusion reaction. A detailed study of the dependence of the 233 U production characteristics as well as energy cost of fissile fuel production of such systems on parameters such as plasma energy gain Q, blanket neutron multiplication has been carried out. The growth rate dynamics of the symbiotic combination of 233 U generating fusion breeders with PHWRs operating on the Th/U-233 cycle in the so called near-breeder regime has been examined. 95% of the energy generated by PHWRs operating with Th/ 233 U fuel would arise from thorium consumption rather than fission of the initially loaded 233 U. A few sub-engineering breakeven fusion breeders producing U-233 at an energy cost well under 200 MeV per atom are adequate to give the requisite nuclear capacity growth rates in conjunction with such near breeder PHWRs. This corresponds to only a 5% diversion of the grid electrical power for the operation of such fusion breeders. In summary the symbiotic combination of a few fusion breeders with a number of PHWRs gives fresh hopes

  19. Improving 900 MW(e) PWR control rooms

    International Nuclear Information System (INIS)

    Bouat, M.; Marcille, R.

    1983-01-01

    Analyses of the behaviour of operators during operating tests on PWR units and the lessons learned from the TMI-2 accident have demonstrated the need to improve the interface between operators and the facilities they control. To that end, and to complement its establishment of safety panels, Electricite de France (EDF) embarked upon a study on the ''Modification of Control Desks and Boards'' in control rooms. This study, involving twenty-eight 900 MW(e) units, almost all of which are currently in service, began with an ergonomic analysis of control rooms by an external consultant, the ADERSA GERBIOS Association. This analysis was based on interviews with simulator instructors and operators, a study of the operation of the unit, and a general review of previous studies. The analysis began in October 1980 and resulted, in April 1981, in a critical report and a proposal to create a full-scale mock-up of a 900 MW(e) control room. Improvements to this were subsequently proposed, enabling options to be made between, among other things, active overall control panels and function-by-function control panels. Finally, a number of general principles, which largely encompass the operators' suggestions, were defined. The alterations to be made will make it necessary to revamp the control panels completely. The work and tests involved should match the duration of refuelling shut-downs. Audio-visual training programmes are planned (portable model). (author)

  20. Qinshan 300Mwe NPP full scope simulator upgrade

    International Nuclear Information System (INIS)

    Qi Kelin; Li Qing; Liu Wei, Lai Shengyuan

    2006-01-01

    On April 28,2004, RINPO was awarded the project for Qinshan 300Mwe NPP full scope simulator upgrade, the SAT (site acceptance test) was completed on June 30 2005 and the simulator put into operator training again. Scope of upgrade includes: computer system (DGI server and workstations) all replaced by microcomputers; G2 I/O controllers all replaced by RTP EIOBC; Unix-based simulation support environment replaced by RINPO's PC-based simulation environment RINSIMTM, Instructor software replaced by RINPO's PC-based instructor software with function and diagram redesigned; DEH, Feed-water control and some other digital control systems redeveloped to follow NPP modifications; desk-top simulator with soft panel control room developed as byproduct; most of the models not changed but it is planned the reactor core and PPC model will be upgraded in near future. SAT of upgrade demonstrates that the performance of the simulator much improved after the upgrade. (author)

  1. Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

    International Nuclear Information System (INIS)

    Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

    1998-07-01

    At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

  2. MOX fuel for Indian nuclear power programme

    International Nuclear Information System (INIS)

    Kamath, H.S.; Anantharaman, K.; Purushotham, D.S.C.

    2000-01-01

    A sound energy policy and a sound environmental policy calls for utilisation of plutonium (Pu) in nuclear power reactors. The paper discusses the use of Pu in the form of mixed oxide (MOX) fuel in two Indian boiling water reactors (BWRs) at Tarapur. An industrial scale MOX fuel fabrication plant is presently operational at Tarapur which is capable of manufacturing MOX fuels for BWRs and in future for PHWRs. The plant can also manufacture mixed oxide fuel for prototype fast breeder reactor (PFBR) and development work in this regard has already started. The paper describes the MOX fuel manufacturing technology and quality control techniques presently in use at the plant. The irradiation experience of the lead MOX assemblies in BWRs is also briefly discussed. The key areas of interest for future developments in MOX fuel fabrication technology and Pu utilisation are identified. (author)

  3. 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

  4. Pramana – Journal of Physics | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Abstract. The radiation damage in the structural materials of a 500 MWe Indian prototype fast breeder reactor (PFBR) is re-assessed by computing the neutron displacement per atom (dpa) cross-sections from the recent nuclear data library evaluated by the USA,ENDF/B-VII.1, wherein revisions were taken place in the new ...

  5. Probabilistic analysis of 900 MWe PWR. Shutdown technical specifications

    International Nuclear Information System (INIS)

    Mattei, J.M.; Bars, G.

    1987-11-01

    During annual shutdown, preventive maintenance and modifications which are made on PWRs cause scheduled unavailabilities of equipment or systems which might harm the safety of the installation, in spite of the low level of decay heat during this period. The pumps in the auxiliary feedwater system, component cooling water system, service water system, the water injection arrays (LPIS, HPIS, CVCS), and the containment spray system may have scheduled unavailability, as well as the power supply of the electricity boards. The EDF utility is aware of the risks related to these situations for which accident procedures have been set up and hence has proposed limiting downtime for this equipment during the shutdown period, through technical specifications. The project defines the equipment required to ensure the functions important for safety during the various shutdown phases (criticality, water inventory, evacuation of decay heat, containment). In order to be able to judge the acceptability of these specifications, the IPSN, the technical support of the Service Central de Surete des Installations Nucleaires, has used probabilistic methodology to analyse the impact on the core melt probability of these specifications, for a French 900 MWe PWR

  6. Study to use graded cobalt adjuster in 540 MWe PHWR

    International Nuclear Information System (INIS)

    Raj, Manish; Fernando, M.P.S.; Pradhan, A.S.; Kumar, A.N.

    2007-01-01

    Full text: There are 17 adjusters in 540 MWe PHWR, which are essentially provided for xenon override function. They also provide flux flattening being in the central region of the reactor core. The present design of adjusters consists of stainless steel tube. The adjuster rods are grouped into 8 banks for movement. Since adjusters are normally fully inserted during reactor operation, they are best suited for production of cobalt 60. The nickel-plated cobalt in the form of either slugs or pellet are used for the design of cobalt pencils. The number of pencils can be varied to optimize the reactivity load and cobalt 60 production requirement. The worth and activity of cobalt adjusters have been worked out considering different pin configuration for the adjuster assembly. To start with we have assumed all adjusters throughout its length are of the same configuration. The flux depression factors within the cobalt pencils have been considered in the estimations of the specific and total cobalt 60 activities. The option of using graded cobalt adjusters, where different pin configuration along the length is considered for better flux flattening

  7. Thermoeconomic Modeling and Parametric Study of Hybrid Solid Oxide Fuel Cell â Gas Turbine â Steam Turbine Power Plants Ranging from 1.5 MWe to 10 MWe

    OpenAIRE

    Arsalis, Alexandros

    2007-01-01

    Detailed thermodynamic, kinetic, geometric, and cost models are developed, implemented, and validated for the synthesis/design and operational analysis of hybrid solid oxide fuel cell (SOFC) â gas turbine (GT) â steam turbine (ST) systems ranging in size from 1.5 MWe to 10 MWe. The fuel cell model used in this thesis is based on a tubular Siemens-Westinghouse-type SOFC, which is integrated with a gas turbine and a heat recovery steam generator (HRSG) integrated in turn with a steam turbi...

  8. 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

  9. Some failures of diesel-generators during commissioning tests of 1300 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Colas, A. F. [Commissariat a l' Energie Atomique, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, CEA/IPSN, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, B.P. No. 6, 92260 Fontenay-aux-Roses (France); Morzelle, C. [Service Etudes et Projets Thermiques et Nucleaires, EdF Lyon (France)

    1986-02-15

    During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: - Alarm sensors on fuel, lubricating, cooling circuits. - Injection pumps and speed governors. - Fuel delivery. - Vibrations of fuel and lubrication lines. This paper will try to show how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects. (authors)

  10. Some failures of diesel-generators during commissioning tests of 1300 MWe PWR

    International Nuclear Information System (INIS)

    Colas, A.F.; Morzelle, C.

    1986-01-01

    During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: - Alarm sensors on fuel, lubricating, cooling circuits. - Injection pumps and speed governors. - Fuel delivery. - Vibrations of fuel and lubrication lines. This paper will try to show how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects. (authors)

  11. Some failures of diesel generators during commissioning tests of 1300 MWe PWR

    International Nuclear Information System (INIS)

    Colas, A.F.; Morzelle, C.

    1985-10-01

    During commissioning tests of the French 1300 MWe units, which are equipped with different diesel generator from the 900 MWe units, some devices and components failures were experienced. These components include: Alarm sensors on fuel, lubricating, cooling circuits; Injection pumps and speed governors; Fuel delivery; Vibrations of fuel and lubrication lines. This paper shows how and when the above elements can affect the reliability of Diesel-generator units and how commissioning tests should show the defects

  12. IRIS-50. A 50 MWe advanced PWR design for smaller, regional grids and specialized applications

    International Nuclear Information System (INIS)

    Petrovic, Bojan; Carelli, Mario; Conway, Larry; Hundal, Rolv; Barbaso, Enrico; Gamba, Federica; Centofante, Mario

    2009-01-01

    IRIS is an advanced, medium-power (1000 MWt or ∼335 MWe) advanced PWR design of integral configuration, that has gained wide recognition due to its innovative 'safety-by-design' safety approach. In spite of its smaller size compared to large monolithic nuclear power plants, it is economically competitive due to its simplicity and advantages of modular deployment. However, the optimum power level for a class of specific applications (e.g., power generation in small regional isolated grids; water desalination and biodiesel production at remote locations; autonomous power source for special applications, etc.) may be even lower, of the order of tens rather than hundreds of MWe. The simple and robust IRIS 335 MWe design provides a solid basis for establishing a 20-100 MWe design, utilizing the same safety and economics principles, so that it will retain economic attractiveness compared to other alternatives of the same power level. A conceptual 50 MWe design, IRIS-50, was initially developed and then assessed in a 2001 report to the US Congress on small and medium reactors, as a design mature enough to have deployment potential within a decade. In the meantime, while the main efforts have focused on the 335 MWe design completion and licensing, parallel efforts have progressed toward the preliminary design of IRIS-50. This paper summarizes the main IRIS-50 features and presents an update on its design status. (author)

  13. 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

  14. Nuclear data needs of Indian nuclear program

    International Nuclear Information System (INIS)

    Fernando, M.P.S.

    2015-01-01

    Currently 17 Pressurised Heavy water Reactors (PHWRs), 2 Boiling water reactors (BWRs) and 1 Pressurised water reactor (PWR) are being operated for power production by Nuclear Power Corporation India Limited (NPCIL). For PHWRs, different types of fuel bundles are simulated by the integral transport theory code, CLUB using a combination of collision probability method and interface current technique and employing IAEA supplied 69 /172 group WIMS cross section library based on ENDF-BVI, BVII. Ring power factors are calculated at different burnups and are used to estimate linear heat rating. The two group neutron cross sections of different type of lattices at different core irradiations are also generated by lattice code CLUB. Wherever reactivity devices are present, supercell approach is adopted and the suitable incremental absorption cross sections are obtained using BOXER which is based on 3-D integral transport theory considering two neutron energy groups. Using the appropriate properties for normal lattices and ones affected by reactivity devices, fuel management and core follow up studies are carried out using 3-D diffusion theory based TRIVENI code. The KAPS-1 power rise transient on March 10, 2004 brought to focus the importance of accurate nuclear data for reactor physics estimation in Indian PHWRs. With IAEA supplied libraries in WIMS format we could satisfactorily resolve the rate of power increase. Stability analysis and sensitivity analysis was carried out for different incore burnup situations resulting from peak flux operation. The quantification of output uncertainties is necessary to adequately establish safety margins of nuclear facilities. The uncertainties in the integral parameters such as reactivity worth and coefficients due to cross section can be assessed using cross section covariance data produced directly from the uncertainties of measurements. Covariance data processing codes and sensitivity analysis tools have to be developed. The part

  15. Sizing of ion exchange column for PHT purification system of 500 MWe Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Goswami, S; Sharma, A K; Bapat, C N; Sharma, V K [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Strict chemistry control is required on the heavy water coolant in the primary loop of the PHWR to ensure stability of oxide corrosion film and to prevent corrosion build up on the surfaces. This is achieved by maintaining PD value of PHT coolant at about 10.45 and sp. conductivity between 20 to 40 mhos/cm by ion exchange column and filtration of coolant. This paper deals with the sizing and selection of bed height of ninth column from different methods. (author). 7 refs., 6 figs.

  16. Upgradation of design features of primary coolant pumps of Indian 220 MWe PHWR

    International Nuclear Information System (INIS)

    Sharma, S.S.; Mhetre, S.G.; Manna, M.M.

    1994-01-01

    Evolution in the design features of Primary Coolant Pump (PCP) had started in fifties for catering to stringent specification requirements of reactor coolant systems of larger capacity reactors of various kinds. Primary coolant pumps of PWR and PHWR are employed for circulating radioactive, pressurized hot water in a circuit consisting of reactor (heat source) and steam generator (heat sink). As primary coolant pump capacity decides the station capacity, larger capacity primary coolant pumps have been evolved. Since primary coolant pump pressure containing parts are part of Primary Heat Transport system envelope, the parts are designed, manufactured, inspected and tested in accordance with the applicable system guidelines. Flywheel is mounted on the motor shaft for increasing mass moment of inertia of pump motor rotor to meet the coast down requirements of reactor cooling system under Class-IV electrical power supply failure. Due to limited accessibility of the PCP (PCP installed in shut down accessible area), quick maintenance, condition monitoring, reliable shaft seal system/bearing system aspects have been of great concern to reactor owners and pump manufacturers. In this paper upgradation of design features of RAPS, MAPS and NAPS primary coolant pumps have been covered. (author). 4 figs., 1 tab

  17. Obligations and characteristics applicable to the French unit of the 1400 MWe series. Adaptation to the 900 and 1300 MWe series

    International Nuclear Information System (INIS)

    Conte, M.

    1985-10-01

    This report presents the directives concerning the obligations and the main characteristics of the nuclear PWR units of 1400 MWe, notified Electricite de France on the 06th of October 1983 by the Industry and Research Department. They reflect the concept of defence in depth [fr

  18. Elasto-plastic finite element analysis of axial surface crack in PHT piping of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Chawla, D.S.; Bhate, S.R.; Kushwaha, H.S.; Mahajan, S.C.

    1994-01-01

    The leak before break (LBB) approach in nuclear piping design envisages demonstrating that the pressurized pipe with a postulated flaw will leak at a detectable rate leading to corrective action well before catastrophic rupture would occur. This requires analysis of cracked pipe to study the crack growth and its stability. This report presents the behaviour of a surface crack in the wall of a thick primary heat transport (PHT) pipe of 500 MWe Indian PHWR. The line spring model (LSM) finite element is used to model the flawed pipe geometry. The variation of crack driving force (J-integral) across the crack front has been presented. The influence of crack geometry factors such as depth, shape, aspect ratio, and loading on peak values of J-integral as well as crack mouth opening displacement has been studied. Several crack shapes have been used to study the shape influence. The results are presented in dimensionless form so as to widen their applicability. The accuracy of the results is validated by comparison with results available in open literature. (author). 47 refs., 8 figs

  19. Comparison of safety margins for leak-before-break assessment of 500 MWe PHWR straight pipes: using contemporary techniques

    International Nuclear Information System (INIS)

    Rastogi, Rohit; Bhasin, Vivek; Kushwaha, H.S.

    1998-01-01

    The Leak Before Break (LBB) analysis of Primary Heat Transport (PHT) Piping of 500 MWe Indian PHWR is being performed using different well established techniques like R6 method (Nuclear Electric UK) and J-Tearing based methods (USNRC). These methods show that PHT piping has required safety margins and can be qualified for LBB. These analysis also showed that the piping has high fracture toughness and plastic collapse is the dominant mode of failure. To enhance the confidence in the results obtained from the above methods, further studies were done on the PHT piping. Procedures which predicted margins against plastic collapse were used. The analysis procedures used were Modified Limit Load Method, MPA Method (both from Germany), Moments Method (from Italy) and the Z-Factor method given in ASME Boiler and Pressure Vessel Code. The safety margins obtained from these analysis satisfied the LBB requirements. A table was generated which compared the safety margins obtained using all the above mentioned procedures. This report presents the results of this study. (author)

  20. Domestic design and validation of natural circulation steam generator of China 1000 MWe PWR NPP

    International Nuclear Information System (INIS)

    Liu, H.Y.; Wang, X.Y.; Wu, G.; Qin, J.M.; Xiong, Ch.H.; Wang, W.; Chen, J.L.; Cheng, H.P.; Zuo, Ch.P.

    2005-01-01

    In order to meet the requirements of domestic design of China intending built NPP projects, Research Institute of Nuclear Power Operation (RINPO) has achieved design of 1000 MWe NPP steam generator, called RINSG-1000(means 1000MWe SG designed by RINPO), which is based on SG research ,experiments and service experience accumulated by RINPO in more 40 years. Testing validation of two steam generator key technologies, advanced moisture separate device and sludge collector, has been accomplished during the period of 2000 to 2002. This paper describes the design features of RINSG-1000, and provides some validation test results. (authors)

  1. Upgrading of fire safety in Indian nuclear power plants

    International Nuclear Information System (INIS)

    Agarwal, N.K.

    1998-01-01

    Indian nuclear power programme started with the installation of 2 nos. of Boiling Water Reactor (BWR) at Tarapur (TAPS I and II) of 210 MWe each commissioned in the year 1996. The Pressurized Heavy Water Reactor (PHWR) programme in the country started with the installation of 2x220 MWe stations at Rawatbhatta near Kota (RAPS I and II) in the State of Rajasthan in the sixties. At the present moment, the country has 10 stations in operation. Construction is going on for 4 more units of 220 MWe where as work on two more 500 MWe units is going to start soon. Fire safety systems for the earlier units were engineered as per the state-of-art knowledge available then. The need for review of fire protection systems in the Indian nuclear power plants has also been felt since long almost after Brown's Ferry fire in 1975 itself. Task forces consisting of fire experts, systems design engineers, O and M personnel as well as the Fire Protection engineers at the plant were constituted for each plant to review the existing fire safety provisions in details and highlight the upgradation needed for meeting the latest requirements as per the national as well as international practices. The recommendations made by three such task forces for the three plants are proposed to be reviewed in this paper. The paper also highlights the recommendations to be implemented immediately as well as on long-term basis over a period of time

  2. Inaugural address

    International Nuclear Information System (INIS)

    Prasad, A.N.

    1994-01-01

    Along with the R and D activities related to technology upgradation in the existing reactors, considerable efforts are being expended in various units of Department of Atomic Energy for the development of advanced technologies to meet the future requirements of Indian nuclear power programme. The two major activities in this thrust area are development of technologies related to the 500 MWe PHWRs and fast breeder reactors. An overview of the activities in some of the thrust areas is given. (author)

  3. 1300°F 800 MWe USC CFB Boiler Design Study

    Science.gov (United States)

    Robertson, Archie; Goidich, Steve; Fan, Zhen

    Concern about air emissions and the effect on global warming is one of the key factors for developing and implementing new advanced energy production solutions today. One state-of-the-art solution is circulating fluidized bed (CFB) combustion technology combined with a high efficiency once-through steam cycle. Due to this extremely high efficiency, the proven CFB technology offers a good solution for CO2 reduction. Its excellent fuel flexibility further reduces CO2 emissions by co-firing coal with biomass. Development work is under way to offer CFB technology up to 800MWe capacities with ultra-supercritical (USC) steam parameters. In 2009 a 460MWe once-through supercritical (OTSC) CFB boiler designed and constructed by Foster Wheeler will start up. However, scaling up the technology further to 600-800MWe with net efficiency of 45-50% is needed to meet the future requirements of utility operators. To support the move to these larger sizes, an 800MWe CFB boiler conceptual design study was conducted and is reported on herein. The use of USC conditions (˜11 00°F steam) was studied and then the changes, that would enable the unit to generate 1300°F steam, were identified. The study has shown that by using INTREX™ heat exchangers in a unique internal-external solids circulation arrangement, Foster Wheeler's CFB boiler configuration can easily accommodate 1300°F steam and will not require a major increase in heat transfer surface areas.

  4. Concept of voltage monitoring for a nuclear power plant emergency power supply system (PWR 1300 MWe)

    International Nuclear Information System (INIS)

    Andrade, R.B. de

    1988-01-01

    Voltage monitoring concept for a Nuclear Power Plant Emergency Power Supply Systems (PWR 1300 MWe) is described based on the phylosophy adopted for Angra 2 and 3 NPP's. Some suggested setpoints are only guidance values and can be modified during plant commissioning for a better performance of the whole protection system. (author) [pt

  5. Development of advanced nuclear fuels in the Indian context: advantages and challenges

    International Nuclear Information System (INIS)

    Ganesan, V.

    2012-01-01

    The ever increasing demand on power requirement in the country has opened up need for exploring use of nuclear fuels that could meet such demands. This makes the mission of the department to shift from the first stage of nuclear programme employing natural uranium in PHWRs to the second stage of deploying a large number of fast reactors with plutonium based fuels capable of realising high breeding ratios in addition to energy production. The transition to fast reactors with advanced fuels, capable of higher breeding ratio, opens up a number of scientific and technological challenges in design and operation of such fast reactors. In the Indian context, after successful demonstration of natural uranium based PHWRs, the performance of U-Pu based carbide fuel, as a unique experience in the world, has been demonstrated in FBTR at Kalpakkam. This paper deals with the performance of carbide fuel in FBTR and the programme on development of metallic fuels with appreciably high breeding ratio that would result in considerable reduction in doubling time thereby addressing the increasing demands of power production as well as pave way for introduction of a large number of such fast reactors to provide energy security to the country. The advantages of introduction of metallic fuels as well as the scientific and technological challenges to be faced in doing so and the ongoing efforts towards metallic fuel development are also described in the paper. (author)

  6. Qualification of PHT piping of Indian 500 MW PHWR for LBB, using R-6 method

    International Nuclear Information System (INIS)

    Rastogi, Rohit; Bhasin, V.; Kushwaha, H.S.

    1997-01-01

    This document discusses the qualification of straight pipe portion of the primary heat transport (PHT) piping of Indian 500 MWe pressurised heavy water reactor (PHWR) for leak before break (LBB). The evaluation is done using R-6 [1] method. The results presented here are: the safety margins which exist on straight pipe components of main PHT piping of 500 MWe, under leakage size crack (LSC) and design basis accident loads; the sensitivity of safety margins with respect to different analysis parameters and the qualification of PHT piping for LBB based on criterion given by NUREG-1061 [2] and TECDOC-774 [3]. (author)

  7. Design and development of improved ballscrew and control circuit for reactivity mechanisms of 220 MWe PHWR operating stations

    International Nuclear Information System (INIS)

    Jain, A.K.; Rama Mohan, N.; Mathew, Jimmy; Mathur, M.K.; Roy, S.; Ingle, V.J.; Ghoshal, B.; Ashok Kumar, B.; Patil, D.C.; Dwivedi, K.P.; Bhambra, H.S.

    2006-01-01

    There has been persistent failure of Ballscrews used for Reactivity Mechanism in standardised 220 MWe PHWR units. The detailed review of failures indicated that on one hand the number of demands for operation of Absorber Rod and Regulating Rod had increased due to use of digital circuit in the drive control system as compared analog circuits used earlier. On the other hand, the existing design of ballscrew had some inherent weaknesses to withstand the loads generated during starting and stopping of the regulating rods. To solve these problems two-pronged approach was adopted. The control problem was traced to overshooting of the servomotor of Absorber Rod and Regulating Rod to the full speed at the time of starting and thereafter, settling to the required speed. This sudden overshooting produces a jerk in the drive mechanism. A modified circuit has been evolved to solve this problem. Also, Changing the dead band and gain of control circuits have reduced the number of rod movements. A 'new design' of Ballscrew assembly was finalised by NPCIL with a view to withstand the severe loads generated during starting and stopping of the regulating rods and to achieve enhanced service life under water-lubrication condition. Based on this design, prototype assemblies were successfully manufactured by two Indian manufacturers. The design was cleared for manufacturing of the bulk production of Ballscrew assemblies after evaluation of its performance during rigorous 'Acceptance Testing'. Two ballscrews of new design were installed in the KGS-1 reactor and are operating since July 2005. This paper covers operational feedback including ballscrew failures in various units, Design/Development of Modified Reactivity Mechanism Ballscrews and Control Circuit based on analysis of underlying causes of failures and feedback on performance of new design. (author)

  8. Main problems experienced on diesel generators of French 900 MWe operating units

    Energy Technology Data Exchange (ETDEWEB)

    Dredemis, Geoffroy; Jude, Francois [Commissariat a l' Energie Atomique, centre d' Etudes Nucleaires de Fontenay-aux-Roses, Institut de Protection et Surete Nucleaire, Departement d' Analyse de Surete, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

    1986-02-15

    Each unit of all the French nuclear power plant is equipped with two diesel emergency generator sets., For the totality of standards PWRs of 900 MWe, they are identical. We present in this communication the most significative failures met with diesel engines on operating units, such as rupture of fuel injection pipes, breaking of the connecting rods, and cylinder lubrication failures. All these incidents, which affected the emergency power sources of concerned units, had generic characteristics. In view of their potential consequences, it was proceeded in each case to an immediate control of the components concerned of all PWR 900 MWe diesel engines. At the same time, studies were started as to what modifications would permit to solve rapidly each one of the problems met with. (authors)

  9. Improvement of Candu-1000 MW(e) power cycle by moderator heat recovery

    International Nuclear Information System (INIS)

    Fath, H.E.S.

    1988-01-01

    Four different moderator heat recovery circuits are proposed for CANDU-1000 MW(e) reactors. The proposed circuits utilize all, or part, of the 155 MW(th) moderator heat load (at 70 0 C moderator outlet temperature from calandria) to the first stage of the feed water heating system. An economics study was carried out and indicated that the direct circulation of feed water through the moderator heat exchanger (with full heat recovery) is the most economical scheme. For this scheme the saved steam from the turbine extraction was found to produce additional electric power of 8 MW(e). This additional power represents a 0.7% increase in the plants nominal electric output. The outstanding features and advantages of the selected scheme are also presented. (author)

  10. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant Conceptual Design Engineering Report (CDER)

    Science.gov (United States)

    1981-01-01

    The reference conceptual design of the magnetohydrodynamic (MHD) Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD, is summarized. Main elements of the design, systems, and plant facilities are illustrated. System design descriptions are included for closed cycle cooling water, industrial gas systems, fuel oil, boiler flue gas, coal management, seed management, slag management, plant industrial waste, fire service water, oxidant supply, MHD power ventilating

  11. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Design Requirements Document (DRD)

    Science.gov (United States)

    Rigo, H. S.; Bercaw, R. W.; Burkhart, J. A.; Mroz, T. S.; Bents, D. J.; Hatch, A. M.

    1981-01-01

    A description and the design requirements for the 200 MWe (nominal) net output MHD Engineering Test Facility (ETF) Conceptual Design, are presented. Performance requirements for the plant are identified and process conditions are indicated at interface stations between the major systems comprising the plant. Also included are the description, functions, interfaces and requirements for each of these major systems. The lastest information (1980-1981) from the MHD technology program are integrated with elements of a conventional steam electric power generating plant.

  12. Secondary cycle water chemistry for 500 MWe pressurised heavy water reactor (PHWR) plant: a case study

    International Nuclear Information System (INIS)

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

    1995-01-01

    In turbine and secondary cycle system of 500 MWe PHWR, chemistry of steam and water is controlled in secondary cycle for prevention of corrosion in steam generators (SGs), feedwater system and steam system, scale and deposit formation on heat transfer surfaces and carry-over of solids by steam and deposition on steam turbine blades. Water chemistry of secondary side of SGs and turbine cycle is discussed. (author). 8 refs., 2 tabs., 1 fig

  13. Role of Fugen-HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, S.

    1982-01-01

    Fugen, a 165 MWe prototype of a heavy water moderated boiling light water cooled reactor; has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work of the 600 MWe demonstration plant has been carried out since 1973. Important system and components, such as pressure tube assemblies, control rod drive mechanism, etc., are essentially the same as those of Fugen. Some modifications, however, are made especially from the stand point of experiences In the Fugen-HWR, plutonium and uranium would be effectively used; and plutonium could make the coolant void reactivity more negative which would give good results in increasing the reactor stability and safety. On the other hand, nuclear power plants are mainly consisted of LWRs in Japan. Considering the above situations, the Fugen-HWR, coupled with LWRs, is now considered in Japan to contribute to our energy security by using plutonium and depleted uranium extracted from spent fuels of LWRs: thereby reducing the demands On August 4, 1981, the ad hoc committee on the 600 MWe demonstration Fugen-HWR submitted the final report to the Japan AEC, after having had discussions and evaluations. In the report, the ad hoc committee recommended to build the 600 MWE demonstration plant with appropriate supports of the Government. The Japan AEC will be expected to make her decision on the program in the near future. As for the reactor safety R and C, development has been stressed on coolant leak detectors and ECCS performances or Since 1965, many development works have been done for mixed oxide fuel assemblies, both for establishment of the fabrication technology and for clarification of irradiation performances. 196 mixed oxide fuel assemblies have been manufactured for Fugen. 168 of them were loaded and 92 were withdrawn. No fuel has been failured yet. (author)

  14. A review of start-up operations on the first units of the 1300 MWe generation

    International Nuclear Information System (INIS)

    Meclot, B.; Lemagny Boc Lonlaygue, C.; Lavogiez, M.

    1986-01-01

    This paper describes and offers comments on the different phases of start-up on power stations of the P4 series. Then, one reviews incidents which occurred in the course of these start-up phases and, having highlighted the lessons to be learnt from the commissioning of these power stations, goes on to make a comparative study of 1300 and 900 MWe availability in the initial year of operation [fr

  15. Evolution of the on-site electric power sources on French 900 MWe PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Bera, Jean [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, Departement d' Analyse de Surete, Service d' Analyse Fonctionnelle, Institut de Protection et Surete Nucleaire, B.P. No. 6, 92260 Fontenay-aux-Roses (France)

    1986-02-15

    Additional means have been provided on the French 900 MWe PWRs to improve safety if both the off-site and on-site Power sources are lost, namely: - a primary pump seal water injection device, one for two units; - a gas turbine generator for each site; - supplying any failing unit with electric power from a house load operating unit; - supplying a unit from a diesel generator of another unit. (author)

  16. Study on evaluating the reactivity worth of the control rods of the PWR 900 MWe

    International Nuclear Information System (INIS)

    Phan Quoc Vuong; Tran Vinh Thanh; Tran Viet Phu

    2015-01-01

    Control rods of a nuclear reactor are divided into two groups: shut down and power control. Reactivity worth of the control rods depends nonlinearly on the rods' compositions and positions where the rods are inserted into the core. Therefore, calculation of control rod worth is of high important. In this study, we calculated the reactivity worth of the power control rod bank of the Mitsubishi PWR 900 MWe. The results are integral and differential worth calibration of the control rods. (author)

  17. Reliability investigation for the ECC subsystem of a 1300 MWe-PWR

    International Nuclear Information System (INIS)

    Lalovic, M.

    1983-01-01

    In this study, a fault-tree analysis is used for reliability investigation of Emergency Core Cooling Sub-system of a 1300 MWe pressurised water reactor. Basic assumptions of the study are large break in the reactor coolant system and independence of the pseudo-components. Relatively high non-availability of the sub-system was calculated. Critical component and minimum cut set are determined. (author)

  18. Fire probability safety analysis in France for 900 MWe nuclear power plants

    International Nuclear Information System (INIS)

    Bertrand, R.; Bonneval, F.; Mattei, J.M.

    2000-01-01

    This paper describes the methodology implemented by the Institute for Nuclear Safety and Protection (IPSN) to carry out the Fire Probabilistic Safety Assessment (Fire PSA) for French 900 MWe pressurised water reactors. The initial results obtained are presented. Additional research and development activities are indicated which IPSN carried out or decided to perform in order to reduce the amount of uncertainty associated with the data or to confirm hypotheses that can impact significantly the study results. (orig.) [de

  19. Role of Fugen HWR in Japan and design of a 600 MWe demonstration reactor

    International Nuclear Information System (INIS)

    Sawai, Sadamu.

    1982-03-01

    Fugen, a 165 MWe prototype of a heavy water-moderated, boiling light water-cooled reactor, has been in commercial operation since March 20, 1979. In parallel with the Fugen project, the design work for a 600 MWe demonstration plant has been carried out since 1973. The important systems and components, such as pressure tube assemblies and control rod drive mechanism, are essentially the same as those of Fugen. However, some modification is made owing to the experience obtained in Fugen and LWrs. In the HWR Fugen, plutonium and uranium are effectively used, and plutonium makes the coolant void reactivity more negative, which results in the increase of the stability and safety of the reactor. On August 4, 1981, the ad hoc committee submitted the final report to the Japanese Atomic Energy Commission, in which the construction of a 600 MWe demonstration plant was recommended. As for the research and development on reactor safety, coolant leak detectors, the performance of ECCS, and safety design codes are enumerated. Since 1965, mixed oxide fuel has been developed, and 168 fuel assemblies were loaded in Fugen, but failure did not occur. (Kako, I.)

  20. Capital cost: high and low sulfur coal plants-1200 MWe. [High sulfur coal

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This Commercial Electric Power Cost Study for 1200 MWe (Nominal) high and low sulfur coal plants consists of three volumes. The high sulfur coal plant is described in Volumes I and II, while Volume III describes the low sulfur coal plant. The design basis and cost estimate for the 1232 MWe high sulfur coal plant is presented in Volume I, and the drawings, equipment list and site description are contained in Volume II. The reference design includes a lime flue gas desulfurization system. A regenerative sulfur dioxide removal system using magnesium oxide is also presented as an alternate in Section 7 Volume II. The design basis, drawings and summary cost estimate for a 1243 MWe low sulfur coal plant are presented in Volume III. This information was developed by redesigning the high sulfur coal plant for burning low sulfur sub-bituminous coal. These coal plants utilize a mechanical draft (wet) cooling tower system for condenser heat removal. Costs of alternate cooling systems are provided in Report No. 7 in this series of studies of costs of commercial electrical power plants.

  1. Development of manufacturing process for production of 500 MWe calandria sheets

    International Nuclear Information System (INIS)

    Hariharan, R.; Ramesh, P.; Lakshminarayana, B.; Bhaskara Rao, C.V.; Pande, P.; Agarwala, G.C.

    1992-01-01

    Calandria tubes made of zircaloy-2 are being used as structural components in pressurised heavy water power reactors. The sheets required for producing calandria tube for 235 MWe reactors are being manufactured at Zircaloy Fabrication Plant (ZFP), NFC utilizing a 2 Hi/4 Hi rolling mill procured for the purpose, by carrying out cold rolling process to achieve the required size after hot rolling suitable extruded slabs. Due to limitation of width of the sheet that can be rolled with the mill as well as the size of the slab that can be extruded with the existing press, difficulties arose in producing acceptable full length sheets of size 6600 mm long x 435 mm wide x 1.6 mm thick for manufacturing 500 MWe calandria tube. This paper deals with the details of the process problem resolved. They are: (a)designing of suitable hot and cold rolling pass schedules, (b)selection and standardization of process parameters such as beta quenching, hot rolling and cold rolling, and (c)details of the overall manufacturing process. Due to implementation of above, sheets required for manufacturing 500 MWe calandria tube sheets were successfully rolled. About 40 nos. of acceptable full length sheets have already been manufactured. (author). 1 fig., 3 tabs

  2. Evaluation of the reliability of the protection system of 1300 MWE PWR'S

    International Nuclear Information System (INIS)

    Blin, A.

    1990-01-01

    An assesment of the reliability of the Digital Integrated Protection System (SPIN) of the 1300 MWe type french reactors has been carried out by treating an example: the emergency shutdown, which can be called upon by several initiating events. The whole chain, from sensors to breakers and control rods, is taken into account. The reliability parameters used for the quantification are evaluated essentially from the experience feedback of french reactors. The not wellknown parameters being the common cause failure rates of electronic components and the efficiency rate of the self-tests, the results of the study are then presented in a parametric form, according to these two factors

  3. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1997-01-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enables faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk represented by deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible. (author)

  4. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    Energy Technology Data Exchange (ETDEWEB)

    Peyrouty, P.

    1996-12-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible.

  5. Capture of SO2 by limestone in a 71 MWe pressurized fluidized bed boiler

    Directory of Open Access Journals (Sweden)

    Shimizu Tadaaki

    2003-01-01

    Full Text Available A 71 MWe pressurized fluidized bed coal combustor was operated. A wide variety of coals were burnt under fly ash recycle conditions. Limestone was fed to the combustor as bed material as well as sorbent. The emission of SO^ and limestone attrition rate were measured. A simple mathematical model of SO? capture by limestone with intermittent solid attrition was applied to the analysis of the present experimental results. Except for high sulfur fuel, the results of the present model agreed with the experimental results.

  6. Design of a 2.5MW(e) biomass gasification power generation module

    Energy Technology Data Exchange (ETDEWEB)

    McLellan, R.

    2000-07-01

    The purpose of this contract was to produce a detailed process and mechanical design of a gasification and gas clean up system for a 2.5MW(e) power generation module based on the generation of electrical power from a wood chip feed stock. The design is to enable the detailed economic evaluation of the process and to verify the technical performance data provided by the pilot plant programme. Detailed process and equipment design also assists in the speed at which the technology can be implemented into a demonstration project. (author)

  7. Improved identification to prevent transposition during operation of 900 MWe PWR reactors

    International Nuclear Information System (INIS)

    Leckner, J.M.; Dien, Y.; Cernes, A.

    1986-04-01

    Detailed human factors analysis of 900 MWe PWR control room identification systems was carried out by the Nuclear and Fossil Generation Division of Electricite de France (EDF) consequent to a series of incidents where personnel confused one plant unit, room or piece of equipment for another. Preliminary analysis uncovered coding inadequacies and suggested possible remedies. This data was used to prepare specifications for identification redesign at a pilot plant on which detailed investigations could be carried out. Recommended solutions were submitted to pilot plant operators and their opinion sollicited. Operator recommendations will be tried out on the pilot plant and adopted on a grid-wide basis if trials prove satisfactory

  8. On-site control of 900 and 1300 MWe nuclear reactors control rod assemblies

    International Nuclear Information System (INIS)

    Lacroix, R.; Lebuffe, C.; Bour, D.; Pasquier, T.

    1990-01-01

    To measure the external wear of clads of the RCCA rodlets in both 900 and 1300 MWe P.W.R., two on site examination tools was developed by FRAGEMA. They have been used in 42 inspections between 1986 and 1989. The examination is performed in two successive phases: - longitudinal detection of wear by eddy currents, - characterization of wear by ultrasonic profilometry. Moreover, at the instance of E.D.F., an equipment is developing by INTERCONTROLE. These measurement tools allow a suitable monitoring system adapted to the phenomenon kinetics [fr

  9. Failures of the thermal barriers of 900 MWe reactor coolant pumps

    International Nuclear Information System (INIS)

    Peyrouty, P.

    1996-01-01

    This report describes the anomalies encountered in the thermal barriers of the reactor coolant pumps in French 900 MWe PWR power stations. In addition to this specific problem, it demonstrates how the fortuitous discovery of a fault during a sampling test enabled faults of a generic nature to be revealed in components which were not subject to periodic inspection, the failure of which could seriously affect safety. This example demonstrates the risk which can be associated with the deterioration in areas which are not examined periodically and for which there are no preceding signs which would make early detection of deterioration possible

  10. 'Kazmer' a complex noise diagnostic system for 1000 MWe PWR WWER type nuclear power units

    International Nuclear Information System (INIS)

    Por, G.

    1992-06-01

    Noise diagnostic systems have previously been developed and installed for the WWER-440 type reactors at the Paks Nuclear Power Plant, Hungary. Based on the experiences, the system has been extended and modified for use in 1000 MWe, WWER-1000 type units. KAZMER consists of three subsystem, the KARD reactor noise diagnostic system, ARGUS vibration monitoring system for rotation machinery, and ALMOS acoustic monitoring system. The installation of the KAZMER system at the Kalinin Nuclear Power Station, Russia, and the first operational experiences are outlined. (R.P.) 15 refs.; 9 figs

  11. The lateral distribution of muons in showers at 40 mwe underground

    International Nuclear Information System (INIS)

    Bergamasco, L.; Castagnoli, C.; Dardo, M.; D'Ettorre Piazzoli, B.; Mannocchi, G.; Picchi, P.; Visentin, R.; Sitte, K.; Freiburg Univ.

    1975-01-01

    The multiplicity distribution of muon showers at 40 mwe underground was studied with a 4 m 2 spark chamber telescope. The observed frequencies deviate systematically from those calculated with the 'standard' lateral distributions of Vernov or of Greisen. Agreement can be attained if an enhancement of the muon component at small shower sizes is assumed, in accordance with the assumptions of a two-component theory of cosmic ray origin. It is improved by introducing an age dependence of the lateral structure function. (orig.) [de

  12. 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

  13. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G.; Zaleski, C.P. [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les

  14. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor; Etudes preliminaires conduisant a un concept de reacteur a neutrons rapides de 1000 MWe

    Energy Technology Data Exchange (ETDEWEB)

    Vendryes, G; Zaleski, C P [Association Euratom-CEA Cadarache (France). Centre d' Etudes Nucleaires

    1964-07-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios ({approx}1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 {delta}k/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [French] Ce rapport presente les etudes qu'il nous a paru important d'aborder dans le cadre du developpement des reacteurs a neutrons rapides. - Il met en evidence l'interet des taux de regeneration internes eleves ({approx}1, 1) pour obtenir une bonne evolution dans le temps de la distribution de puissance et de la reactivite (moins de 0,005 {delta}k/k pour 18 mois). - Il montre la possibilite d'y parvenir tout en applatissant la distribution des fissions, ce qui se traduit par une puissance specifique moyenne plus elevee (gain economique), et donc un temps de doublement plus faible de l'ordte de 10 ans - Il tente de definir un optimum de la puissance specifique valable pour les projets de reacteurs futurs

  15. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.M.

    1991-08-01

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990

  16. Role of pressuriser in enhancing pressure control system capability in primary system of 500 MWe PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Walia, M P.S.; Misri, Vijay; Bapat, C N; Sharma, V K [Nuclear Power Corporation, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    The primary heat transport system of a pressurized heavy water reactor (PHWR) extracts and transports the heat produced in the fuel (located inside coolant channel assemblies) to the steam generators where steam is generated to run the turbo-generator. The heat transport medium (primary coolant) is heavy water which is kept in a pressurized liquid state with the help of a pressure control system. Feed and bleed circuits with associated equipment of PHT main system have traditionally constituted the pressure control system. However, for large size reactors of 500 MWe capacity, a surge tank known as pressurizer was incorporated due to the presence of relatively large inventory in PHT main circuit. The pressurizer acts as a cushion for pressure variations resulting from various transients. This significantly reduces the onerous demand on feed and bleed system, thereby reducing reactor outages on system pressure excursions. The paper describes in detail the pressure control system of 500 MWe PHWR involving pressuriser and feed and bleed system including their functions and instrumentation. The results of mathematical modelling/analysis undertaken to establish the response adequacy of pressure control system, to postulated plant transients vis-a-vis the role of pressurizer are presented. (author). 10 figs.

  17. ASTEC-CATHARE2 benchmarks on French PWR 1300MWe reactors

    International Nuclear Information System (INIS)

    Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

    2009-01-01

    The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe reactors. This PSA-2 is heavily relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, an important series of benchmarks with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe accident scenarios. The present paper details 2 out of the 14 studied scenarios: a 12 inches cold leg Loss of Coolant Accident (LOCA) and a 2 tubes Steam Generator Tube Rupture (SGTR). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and the ASTEC results of the core degradation phase are presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results. (author)

  18. Source terms associated with two severe accident sequences in a 900 MWe PWR

    International Nuclear Information System (INIS)

    Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

    1983-12-01

    Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

  19. Probabilistic safety assessment of French 900 and 1,300 MWe nuclear plants

    International Nuclear Information System (INIS)

    Brisbois, J.; Lanore, J.M.

    1991-01-01

    Although reactor design is mainly governed by deterministic principles in France, the probabilistic approach has been considered an important aid to safety analysis since the early seventies. Various partial probabilistic studies have been performed by Electricite de France, by IPSN and by Framatome, for various types of reactor. In particular, these studies have made it possible to assess the reliability and availability of nuclear power plants safety systems as well as the probability of accident scenarios and have helped to define technical specifications (especially, allowed operating times in the event of a partial unavailability of safety systems). Simultaneously, evaluation methods and corresponding software have been widely developed. Besides. EDF has implemented the Systeme de Recueil de Donnees de Fiabilite - SRDF (Reliability Data Collection System) which allows follow-up of equipment behaviour on all the operating units, and has led to a particularly representative data base. In 1982 the decision was taken at IPSN to carry out a complete PSA for a standard reactor of the 900 MWe type, and in 1986 EDF launched an equivalent study on a 1,300 MWe reactor, taking Unit 3 Paluel as reference. These PSAs were terminated in the course of the first quarter of 1990. (author)

  20. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  1. 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

  2. Neutron radiation damage studies in the structural materials of a 500 MWe fast breeder reactor using DPA cross-sections from ENDF / B-VII.1

    Science.gov (United States)

    Saha, Uttiyoarnab; Devan, K.; Bachchan, Abhitab; Pandikumar, G.; Ganesan, S.

    2018-04-01

    The radiation damage in the structural materials of a 500 MWe Indian prototype fast breeder reactor (PFBR) is re-assessed by computing the neutron displacement per atom (dpa) cross-sections from the recent nuclear data library evaluated by the USA, ENDF / B-VII.1, wherein revisions were taken place in the new evaluations of basic nuclear data because of using the state-of-the-art neutron cross-section experiments, nuclear model-based predictions and modern data evaluation techniques. An indigenous computer code, computation of radiation damage (CRaD), is developed at our centre to compute primary-knock-on atom (PKA) spectra and displacement cross-sections of materials both in point-wise and any chosen group structure from the evaluated nuclear data libraries. The new radiation damage model, athermal recombination-corrected displacement per atom (arc-dpa), developed based on molecular dynamics simulations is also incorporated in our study. This work is the result of our earlier initiatives to overcome some of the limitations experienced while using codes like RECOIL, SPECTER and NJOY 2016, to estimate radiation damage. Agreement of CRaD results with other codes and ASTM standard for Fe dpa cross-section is found good. The present estimate of total dpa in D-9 steel of PFBR necessitates renormalisation of experimental correlations of dpa and radiation damage to ensure consistency of damage prediction with ENDF / B-VII.1 library.

  3. Adding a much needed 300 MWe at South Africa's Arnot coal fired power plant

    Energy Technology Data Exchange (ETDEWEB)

    Rich, G. [Alstom, Rugby (United Kingdom)

    2008-12-15

    As power stations built in the last thirty years approach the end of their design life, and the cost of new capacity continues to increase, along with demands for improved efficiency and lower emissions, an integrated approach to retrofit looks increasingly compelling. The ambitious upgrade project currently underway at the Arnot coal fired plant in South Africa, which will result in an update from 6 x 350 MWe to 6 x 400 MWe and a life extension of 20 years, illustrates the benefits. 2 figs.

  4. Evaluation of feed and bleed cooling mode in case of total loss of feedwater on 900 MWe PWR

    International Nuclear Information System (INIS)

    Champ, M.; Cornille, Y.

    1989-07-01

    The physical studies carried out with the CATHARE code to assess the feed and bleed procedure developed in order to cope with the total loss of feed water on a 900 MWe PWR are presented. These studies allowed the definition of the maximum delays of intervention which would prevent the core from uncovering. Different cases of equipment availability are considered. The data generated will be used in the 900 MWe Probabilistic Safety Assessment which is under way at the Institut de Protection et de Surete Nucleaire

  5. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-04-15

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10{sup -6} per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective

  6. A probabilistic safety assessment of the standard French 900MWe pressurized water reactor. Main report

    International Nuclear Information System (INIS)

    1990-04-01

    To situate the probabilistic safety assessment of standardized 900 MWe units made by the Institute for Nuclear Safety and Protection (IPSN), it is necessary to consider the importance and possible utilization of a study of this type. At the present time, the safety of nuclear installations essentially depends on the application of the defence in-depth approach. The design arrangements adopted are justified by the operating organization on the basis of deterministic studies of a limited number of conventional situations with corresponding safety margins. These conventional situations are grouped in categories by frequency, it being accepted that the greater the consequences the lesser the frequency must be. However in the framework of the analysis performed under the control of the French safety authority, the importance was rapidly recognized of setting an overall reference objective. By 1977, on the occasion of appraisal of the fundamental safety options of the standardized 1300 MWe units, the Central Service for the Safety of Nuclear Installations (SCSIN) set the following global probabilistic objective: 'Generally speaking, the design of installations including a pressurized water nuclear reactor must be such that the global probability of the nuclear unit being the origin of unacceptable consequences does not exceed 10 -6 per year...' Probabilistic analyses making reference to this global objective gradually began to supplement the deterministic approach, both for examining external hazards to be considered in the design basis and for examining the possible need for additional means of countering the failure of doubled systems in application of the deterministic single-failure criterion. A new step has been taken in France by carrying out two level 1 probabilistic safety assessments (calculation of the annual probability of core meltdown), one for the 900 MWe series by the IPSN and the other for the 1300 MWe series by Electricite de France. The objective of

  7. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    The research was useful for in-service performance evaluation, safety assessment, residual life estimation and life extension of nuclear reactors built during stage I i.e., PHWRs and BWRs. It also included developments which would permit rapid expansion of nuclear power initially through fast breeder reactor based on ...

  8. FEM analysis of foundation raft for 500 MWe pressurized heavy water reactor building

    International Nuclear Information System (INIS)

    Kulkarni, N.N.; Goray, J.S.; Joshi, M.H.; Paramasivam, V.

    1989-01-01

    Foundation raft supports the containment structure and internals for 500 MWe PHW reactor building. It also serves as bottom envelop of the containment structure. In view of this, the design of foundation raft assumes great importance. The foundation raft is subjected to various load, most significant of them are dead load of structure, equipment loads transferred through a system of floors, walls and structural steel columns, pressure load during accident conditions, seismic loads, earth pressure, uplift due to buoyancy loads, foundation reaction etc. In order to achieve optimum design, the detailed structural analysis is required to be performed methodically and in most realistic manner. Finite element methods which have come in vogue with the developments in digital computers can be successfully applied in this area. The paper describes the above methods in detail for the analysis of foundation raft for the various load combinations required to be considered for safe and optimum design

  9. Evaluation of full MOX core capability for a 900 MWe PWR

    International Nuclear Information System (INIS)

    Joo, Hyung-Kook; Kim, Young-Jin; Jung, Hyung-Guk; Kim, Young-Il; Sohn, Dong-Seong

    1996-01-01

    Full MOX capability of a PWR core with 900 MWe capacity has been evaluated in view of plutonium consumption and design feasibility as an effective means for spent fuel management. Three full MOX cores have been conceptually designed; for annual cycle, for 18-month cycle, and for 18-month cycle with high moderation lattice. Fissile and total plutonium quantities at discharge are significantly reduced to 60% and 70% respectively of initial value for standard full MOX cores. It is estimated that one full MOX core demands about 1 tonne of plutonium per year to be reloaded, which is equivalent to reprocessing of spent nuclear fuels discharged from five nuclear reactors operating with uranium fuels. With low-leakage loading scheme, a full MOX core with either annual or 18-month cycle can be designed satisfactorily by installing additional rod cluster control system and modifying soluble boron system. Overall high moderation lattice case promises better core nuclear characteristics. (author)

  10. 400-MWe consolidated nuclear steam system (CNSS). 1255 MWt CNSS design/cost update

    International Nuclear Information System (INIS)

    1984-07-01

    Since 1976 Babcock and Wilcox (B and W) has been extensively involved in the development of a medium-sized (1255 MWt/400 MWe) reactor. Under the sponsorship of the U.S. Department of Energy (DOE) and through a contract with Oak Ridge National Laboratories (ORNL), B and W investigated the feasibility of the concept for utility power generation and cogenerated process heat. The potential benefits of the design, called the Consolidated Nuclear Steam System (CNSS), were also identified. This study provides an update of the CNSS design and cost reflecting current regulatory requirements and operating reactor experience. The study was funded by DOE through ORNL and was performed by B and W and UE and C

  11. Recent operating experience during startup testing at latest 1100 MWe BWR-5 nuclear power plants

    International Nuclear Information System (INIS)

    Tanabe, Akira; Tateishi, Mizuo; Kajikawa, Makoto; Hayase, Yuichi.

    1986-01-01

    In June and September 1985, the latest two 1100 Mwe BWR-5 nuclear power plants started commercial operation about ten days earlier than initially expected without any unscheduled shutdown. These latest plants, 2F-3 and K-1, are characterized by an improved core with new 8 x 8 fuel assemblies, highly reliable control systems, advanced control room system and turbine steam full bypass system for full load rejection (2F3). This paper describes the following operating experiences gained during their startup testing. 1) Continuous operation at full load rejection. 2) Stable operation at natural circulating flow condition. 3) 31 and 23 hour short time start up operation. 4) 100-75-100 %, 1-8-1-14 hours daily load following operation. (author)

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  13. Modification to 200 MW(e) CANDU for improved dynamic behaviour

    International Nuclear Information System (INIS)

    Chamany, B.F.; Murthy, L.G.K.; Ray, R.N.

    1976-01-01

    Rajasthan Atomic Power Station is inherently suitable for base load operation. Its control philosophy is based on turbine following the reactor. However, due to load fluctuations and inherent limitation of the control system, there had been considerable number of outages of the station. This limitation is further enhanced by improper choice of the operating pressure range of the boilers. Besides, existing fuel design does not permit thermal cycling and hence there is no use in attempting to make the reactor follow the turbine. Design modifications have been suggested for incorporation in the further 200 MW(e) systems. The method adopted is complete decoupling of the reactor from the load. Dynamic behaviour of the station with the suggested modifications and its comparison with the existing situation has been brought out. (author)

  14. Current status of 700 MWe class PHWR NSSS design and engineering technology

    International Nuclear Information System (INIS)

    Park, Tae Keun; Suh, Sung Ki

    1996-06-01

    The capability of NSSS design and engineering technology of KAERI for 700 MWe class PHWR (CANDU 6) as of 1996 March 30 is comprehensively summarized in this report. The design and engineering capability of KAERI which have been gained during the implementation of Wolsung 2, 3 and 4 project are assessed, and showed with tangible scale. The status of Technology Transfer Materials received from Atomic Energy of Canada Limited under the Technology Transfer Agreement (TTA) which is effective simultaneously to Wolsung 3 and 4 contract, is also given in this report. The division of responsibility (DOR) of KAERI for Wolsung 2 and Wolsung 3 and 4 contract is also given, and expansion of DOR from Wolsung 2 contract to Wolsung 3 and 4 is presented. 3 refs. (Author)

  15. Systems analysis of a 100-MWe modular liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Morris, E.E.; Rhow, S.K.; Switick, D.M.

    1985-01-01

    The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

  16. Control of the flanges of the thermal barriers fitting the 900 MWe PWR primary pumps

    International Nuclear Information System (INIS)

    Cleurennec, M.; Thebault, Y.; Abittan, E.; Pages, C.; Lhote, P.A.; Randrianarivo, L.

    1998-01-01

    During maintenance visit on 93 D type primary pumps of French 900 MWe nuclear units, cracking has been evidenced on the thermal barrier, first on the flange, on the face of connection of the cooling, water coils, and then on the weld between the housing and the flange. Laboratory examinations have exhibited that this cracking is due to a fatigue phenomenon which is initiated on locations where high residual stresses are present. One pump, in service in a plant, has received an instrumentation in order to determine stress cycling. Measurements of temperature on the surface of the metal have shown the presence of thermal cycling due to the thermohydraulic conditions inside the thermal barrier. A non destructive testing method using ultrasounds has been developed in order to asses the magnitude cracking. Corrective and preventive actions have been implemented for repairing and improving thermal barrier when cracking is detected. (authors)

  17. Co-firing straw and coal in a 150-MWe utility boiler: in situ measurements

    DEFF Research Database (Denmark)

    Hansen, P. F.B.; Andersen, Karin Hedebo; Wieck-Hansen, K.

    1998-01-01

    A 2-year demonstration program is carried out by the Danish utility I/S Midtkraft at a 150-MWe PF-boiler unit reconstructed for co-firing straw and coal. As a part of the demonstration program, a comprehensive in situ measurement campaign was conducted during the spring of 1996 in collaboration...... with the Technical University of Denmark. Six sample positions have been established between the upper part of the furnace and the economizer. The campaign included in situ sampling of deposits on water/air-cooled probes, sampling of fly ash, flue gas and gas phase alkali metal compounds, and aerosols as well...... deposition propensities and high temperature corrosion during co-combustion of straw and coal in PF-boilers. Danish full scale results from co-firing straw and coal, the test facility and test program, and the potential theoretical support from the Technical University of Denmark are presented in this paper...

  18. Estimate of man-rem expenditures for a mature CANDU 600 MW(e) station

    International Nuclear Information System (INIS)

    Kuperman, I.

    1978-08-01

    In recent years, man-rem expenditures at operating stations have come under close scrutiny in order to reduce operating personnel dosage. This increased awareness has led to concerted efforts to improve station design and to improve operating procedures to achieve lower man-rem expenditures. This paper is intended to highlight design improvements that have been made in the CANDU 600 MW(e) design and to show how these improvements will reduce man-rem expenditures. Other considerations, such as station decontaminations of the primary heat transport system and the fuelling machines and stricter chemistry control are presently available to help reduce man-rem consumption. Also, station management operating policy should emphasize man-rem awareness. (author)

  19. Ergonomic design of mosaic control panel and standardised control tile configurations for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Ughade, A.V.; Das, R.N.; Ramakrishnan, S.

    1994-01-01

    A review of control rooms of operating nuclear power plants identified many design problems having potential for degrading the performance of operators. Many indications and controls on existing control panels are placed outside the recommended visual and reach envelopes for acceptable operator usage. As a result, the application of human factor principles was found to be needed. This paper describes the design approach for working out the dimensions of main control room panels and console using human engineering principles and recommends the ergonomic dimensions of the main control room panels and console. Further it gives the basis and works out the control tile configurations for 500 MWe PHWR project. It also suggests the use of a full scale mock up for design evaluation and verification. (author). 7 refs., 4 figs

  20. Design Evaluation of UIS and In-vessel Fuel Transfer Machine for a 1200MWe SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Kim, Seok Hoon; Park, Chang Gyu; Lee, Su Yeon

    2008-11-15

    The report describes the structural applicability of the upper internal structure (UIS) and the in-vessel fuel transfer machine for a 1200MWe sodium cooled fast reactor (SFR) of a pool type. In the conceptual design, a two rotating plug type as a refueling system is considered. For the two rotating plug type, the diameters of large and small rotating plugs are determined by 7.2m and 5.6m, respectively. Through the use of an inner fuel transfer machine and the lift change machine with a fixed arm length of 1.10m installed on a small rotating plug, all the core assemblies are accessed within 7mm accuracy. The UIS diameter is determined by 4.7m, which includes the all control drive lines in upper part, the diameter of UIS lower part is restricted by 4.4 m to secure the rotation angle of a refueling machine.

  1. Considerations in providing purification flows for 500 MWe PHWR primary circuits

    International Nuclear Information System (INIS)

    Sharma, A.K.; Goswami, S.; Bapat, C.N.; Sharma, V.K.

    1995-01-01

    The purpose of the purification system is to keep the primary heat transport (PHT) system clean by removing traces of impurities arising due to corrosion of the carbon steel pipes and heat transfer surfaces and erosion/corrosion of valve trims, pipes and mechanical seals or due to presence of soluble or insoluble fission products. These impurities are undesirable because they are usually radioactive, either naturally or through activation by the neutron flux as they are carried by the coolant through the reactor core. The purification system minimizes the probability of generation of radioactive impurities by controlling the chemistry of PHT coolant so that corrosion is minimum. Various considerations for providing the requisite purification flow to fulfill the above functions for a typical 500 MWe PHWR are presented. (author). 4 refs., 2 tabs., 2 figs

  2. The Clean Coal Technology Program 100 MWe demonstration of gas suspension absorption for flue gas desulfurization

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, F.E.; Hedenhag, J.G. [AirPol Inc., Teterboro, NJ (United States); Marchant, S.K.; Pukanic, G.W. [Dept. of Energy, Pittsburgh, PA (United States). Pittsburgh Energy Technology Center; Norwood, V.M.; Burnett, T.A. [Tennessee Valley Authority, Chattanooga, TN (United States)

    1997-12-31

    AirPol Inc., with the cooperation of the Tennessee Valley Authority (TVA) under a Cooperative Agreement with the United States Department of Energy, installed and tested a 10 MWe Gas Suspension Absorption (GSA) Demonstration system at TVA`s Shawnee Fossil Plant near Paducah, Kentucky. This low-cost retrofit project demonstrated that the GSA system can remove more than 90% of the sulfur dioxide from high-sulfur coal-fired flue gas, while achieving a relatively high utilization of reagent lime. This paper presents a detailed technical description of the Clean Coal Technology demonstration project. Test results and data analysis from the preliminary testing, factorial tests, air toxics texts, 28-day continuous demonstration run of GSA/electrostatic precipitator (ESP), and 14-day continuous demonstration run of GSA/pulse jet baghouse (PJBH) are also discussed within this paper.

  3. Concept of voltage and frequency monitoring for a nuclear power plant normal power supply system - PWR 1300 MWe

    International Nuclear Information System (INIS)

    Andrade, R.B. de

    1990-01-01

    Voltage and frequency monitoring concept for a Nuclear Power Plant Normal Power Supply System (PWR 1300 MWe) is described based on the phylosophy adopted for Angra 2 and e NPP's. Some suggested setpoints are only guidance values and can be modified during plant commissioning for a better performance of the whole protection system. (author) [pt

  4. 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)

  5. Indian Legends.

    Science.gov (United States)

    Gurnoe, Katherine J.; Skjervold, Christian, Ed.

    Presenting American Indian legends, this material provides insight into the cultural background of the Dakota, Ojibwa, and Winnebago people. Written in a straightforward manner, each of the eight legends is associated with an Indian group. The legends included here are titled as follows: Minnesota is Minabozho's Land (Ojibwa); How We Got the…

  6. Challenges of atomic energy regulation in Indian context

    International Nuclear Information System (INIS)

    Bajaj, S.S.

    2010-01-01

    Over the years, India has mastered all the stages of the nuclear fuel cycle, which include mining, processing and fabrication of nuclear fuel; design, construction, and operation of nuclear power reactors and research reactors; reprocessing of spent fuel and management of radioactive wastes. Ionising radiation is also used widely in medical, industrial and research areas. Since its inception, Department of Atomic Energy (DAE) was enforcing radiological safety in the country through in-house or ad-hoc committees, till a dedicated regulatory body (AERB) was set up 25 years ago. Today India is operating 19 nuclear power plants with different vintages (2 BWRs and 17 PHWRs) and another 8 (1 PFBR, 5 PHWRs and 2 PWRs) are in various stages of construction. Recently there are new evolutionary reactors (AHWRs) for which design has been completed and are on the threshold for consideration for construction. To match the rapid growth in the need for power India is also about to take up construction of large evolutionary PWRs of foreign design. This variety in the Indian nuclear power programme has come up due to a systematic evaluation and optimisation of the resources and technology available within the country. Added to this is the growing use of radiation in non-power applications. As the safety supervision of this huge programme is the responsibility of AERB, it faces various challenges, like, - Strategies for regulating wide variety of nuclear and radiation facilities with wide dispersal; - Meeting present day expectations with regard to nuclear and radiation safety and nuclear security; - The safety and security of large number of radioactive sources spread over such a vast country and of the associated import/export guidance; - Ensuring safety of old plants by periodic reviews and by prescribing adequate safety upgradation and ageing management programme; -Adaptation of the regulatory system and of regulations to new and foreign design nuclear technologies and

  7. Improved NOx emissions and combustion characteristics for a retrofitted down-fired 300-MWe utility boiler.

    Science.gov (United States)

    Li, Zhengqi; Ren, Feng; Chen, Zhichao; Liu, Guangkui; Xu, Zhenxing

    2010-05-15

    A new technique combining high boiler efficiency and low-NO(x) emissions was employed in a 300MWe down-fired boiler as an economical means to reduce NO(x) emissions in down-fired boilers burning low-volatile coals. Experiments were conducted on this boiler after the retrofit with measurements taken of gas temperature distributions along the primary air and coal mixture flows and in the furnace, furnace temperatures along the main axis and gas concentrations such as O(2), CO and NO(x) in the near-wall region. Data were compared with those obtained before the retrofit and verified that by applying the combined technique, gas temperature distributions in the furnace become more reasonable. Peak temperatures were lowered from the upper furnace to the lower furnace and flame stability was improved. Despite burning low-volatile coals, NO(x) emissions can be lowered by as much as 50% without increasing the levels of unburnt carbon in fly ash and reducing boiler thermal efficiency.

  8. Analysis of the in-vessel phase of SAM strategy for a Korean 1000 MWe PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Sung-Min; Oh, Seung-Jong [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Diab, Aya [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of). Dept. of NPP Engineering; Ain Shams Univ., Cairo (Egypt). Mechanical Power Engineering Dept.

    2017-12-15

    This paper focuses on the in-vessel phase of Severe Accident Management (SAM) strategy for a Korean 1000 MWe Pressurized Water Reactor (PWR) with reference to ROAAM+ framework approach. To apply ROAAM+, it is needed to identify epistemic and aleatory uncertainties. The selected scenario is a station blackout (SBO) and the corresponding SAM strategy is RCS depressurization followed by water injection into the reactor pressure vessel (RPV). The analysis considers the depressurization timing and the flow rate and timing of in-vessel injection for scenario variations. For the phenomenological uncertainties, the core melting and relocation process is considered to be the most important phenomenon in the in-vessel phase of SAM strategy. Accordingly, a sensitivity analysis is carried out to assess the impact of the cut-off porosity below which the flow area of a core node is zero (EPSCUT), and the critical temperature for cladding rupture (TCLMAX) on the core melting and relocation process. In this paper, the SAM strategy for maintaining the integrity of RPV is derived after quantification of the scenario and phenomenological uncertainties.

  9. Final report on the evolution of supporting conditions for the feeders of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Soni, R.S.; Kushawaha, H.S.; Mahajan, S.C.; Kakodkar, A.; Hariprasad, K.

    1994-01-01

    This report deals with the evolution of generic supporting conditions for the feeders of 500 MWe PHWR based on the analysis and qualification of a few representative feeders. There are 196 different feeder pipe configurations for a total of 748 feeders. The present analysis was aimed at evolving a generalised supporting criteria based on the analysis of some representative feeders. The analysis was carried out for various loadings viz. pressure, temperature, dead weight, operating basis earthquake (OBE), safe shutdown earthquake (SSE) and creep loadings. The analysis for OBE and SSE loadings were carried out using response spectrum method. The effect of spacers between various feeders was modelled using higher damping values than those prescribed in ASME code. Based on the above analyses, generic supporting arrangements for the feeders of various groups have been finalized. This report gives details about the mathematical modelling, the analysis approach, the optimised supporting criteria, finalization of grouping and fixing of boundaries between various groups of feeders. (author). 34 refs., 51 figs., 69 tabs

  10. Mathematical modelling of heat absorption capacity of containment spray system in a 700 MWe PHWR

    International Nuclear Information System (INIS)

    Kota, Sampath Bharadwaj; Ali, Seik Mansoor; Balasubramaniyan, V.

    2015-01-01

    This paper presents a mathematical model for estimating the heat removal by containment spray system in the post Loss of Coolant Accident (LOCA) environment. The procedure involves firstly, the calculation of heat removal rates by droplets of spray dispersed in the air-steam mixture by an appropriate direct contact condensation model accounting for the presence of non-condensable gas (air). Parametric influence of droplet size, ambient pressure and temperature on heat flux is brought out. It was found that the heat flux is inversely proportional to the ambient pressure and diameter. A spray module was subsequently developed and incorporated into an in-house containment thermal hydraulics code. The pressure and temperature transients in a 700 MWe PHWR containment building following a Large Break LOCA was obtained using this code. The efficacy of the spray in condensing the steam is shown by comparing the transients with and without the operation of spray system. Parametric studies are also conducted with respect to droplet size and flow rate of water droplet spray. The details of the investigation are presented and discussed in this paper. (author)

  11. Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Research and Development Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, E.S.; Homer, G.B.; Shaber, C.R.; Thurow, T.L.

    1981-11-17

    The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.

  12. Internal Technical Report, Safety Analysis Report 5 MW(e) Raft River Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, E.S.; Homer, G.B.; Spencer, S.G.; Shaber, C.R.

    1980-05-30

    The Raft River Geothermal Site is located in Southern Idaho's Raft River Valley, southwest of Malta, Idaho, in Cassia County. EG and G idaho, Inc., is the DOE's prime contractor for development of the Raft River geothermal field. Contract work has been progressing for several years towards creating a fully integrated utilization of geothermal water. Developmental progress has resulted in the drilling of seven major DOE wells. Four are producing geothermal water from reservoir temperatures measured to approximately 149 C (approximately 300 F). Closed-in well head pressures range from 69 to 102 kPa (100 to 175 psi). Two wells are scheduled for geothermal cold 60 C (140 F) water reinjection. The prime development effort is for a power plant designed to generate electricity using the heat from the geothermal hot water. The plant is designated as the ''5 MW(e) Raft River Research and Development Plant'' project. General site management assigned to EG and G has resulted in planning and development of many parts of the 5 MW program. Support and development activities have included: (1) engineering design, procurement, and construction support; (2) fluid supply and injection facilities, their study, and control; (3) development and installation of transfer piping systems for geothermal water collection and disposal by injection; and (4) heat exchanger fouling tests.

  13. Design of reactor components (non replaceable) of 500 MWe PHWR for enhanced life

    International Nuclear Information System (INIS)

    Dwivedi, K.P.; Seth, V.K.

    1994-01-01

    A nuclear power station is characterised by large initial cost and low operating cost. So a plant which is capable of operating for a longer period of time will be economically more attractive. In the past approach had been to design a nuclear power plant for 30 to 40 years of life time. However, with the improvement in technology and incorporation of redundant and diverse safety features it is now possible to design a nuclear power plant for longer life. Now internationally it is being realised that without sacrificing safety features, plant life should be extended till the cost of maintenance or refurbishment is larger than the cost of the replacement capacity. In order to meet the objective of long life, for the components which cannot be easily replaced the life time of about 100 years is being considered as the design objective. For other items replacement, layout space, shielding, access route and lifting capacity and component design are receiving additional emphasis so as to provide a long total station life time. With the above background, design improvements to enhance the life of reactor components for 500 MWe PHWR namely calandria, end shields and calandria vault liners which cannot be replaced and on which any repair is extremely difficult, have been made. This paper deals with design life of these components and the modifications incorporated in the design. (author). 3 refs., 2 tabs., 3 figs

  14. Interpretation of out of line control rod experiments for 1300 MWE PWR

    Energy Technology Data Exchange (ETDEWEB)

    Leroy, J.L.; Garcia-Fernandez, L.

    1988-01-01

    The present note summarizes the studies we performed recently in order to search a 2D reconstruction procedure for the 1300 MWE PWR power shape, starting from data coming out from thermocouples placed on several fuel assemblies. In classical PWR design, only a few assemblies are equipped with measurement devices, so that it is necessary to interpolate among measure points in order to obtain a complete coverage of the core. A mathematical approach based on the splitting of the power into a reference steady state nominal shape and some ''influence'' and harmonic functions was chosen. The reference steady state power shape, which corresponds to the full power operating mode, is obtained via direct mobile chamber measurements. The perturbations due to the control rod movements are accounted for by specific ''influence'' functions: moreover, harmonics are used to reconstruct the minor effects due to xenon tilts, rod out of line positions and all actual mechanical and thermohydraulic inhomogeneities. The weighting coefficients of the functions are evaluated by a least square method, starting from the distribution of the deviations among the measurements and the reference values.

  15. Stress and fatigue analysis of fuelling machine housing of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Dutta, B.K.; Ramana, W.V.; Kushwaha, H.S.; Kakodkar, A.

    1987-01-01

    One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine rive mechanisms. Some of these holes intersect with each other in the housing end-closers and hence end-closers are reinforced accordingly. This also makes the end-closers nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described. (orig.)

  16. Stress and fatigue analysis of fuelling machine housing of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Dutta, B.K.; Ramana, W.V.; Kushwaha, H.S.; Kakodkar, A.

    1987-01-01

    One of the most appealing features of the Pressurised Heavy Water Reactors is the online refuelling capability. For this a fuelling machine is used. This machine opens a reactor channel by removing a seal plug and a shield plug and then does the necessary fuelling by pushing fuel bundles from a fuel magazine by rams. After necessary fuelling the machine closes the channel automatically. One of the most important parts of the fuelling machine is its pressure housing which becomes a part of the reactor channel during refuelling operation. It houses the fuel magazine, separators and rams. Beside channel pressure and other mechanical loads, the pressure housing experiences thermal transients during refuelling. The housing consists of two cylindrical shells having one end-closer in each. They are connected with each other by a large sized coupling. There are many holes on both the end-closers to accommodate ram movement, separators and magazine drive mechanisms. Some of these holes intersect with each other in the housing end-closures and hence end-closures are reinforced accordingly. This also makes the end-closures nonsymmetric. In the following sections the various analysis done to compute general stress distribution, stress concentration factors near to various holes, temperature transients during refuelling and also allowable fatigue cycles for pressure housing of fuelling machine for the proposed 500 MWe are described

  17. Preliminary analysis of a large 1600 MWe PWR core loaded with 30% MOX fuel

    International Nuclear Information System (INIS)

    Polidoro, Franco; Corsetti, Edoardo; Vimercati, Giuliano

    2011-01-01

    The paper presents a full-core 3-D analysis of the performances of a large 1600 MWe PWR core, loaded with 30% MOX fuel, in accordance with the European Utility Requirements (EUR). These requirements state that the European next generation power plants have to be designed capable to use MOX (UO 2 - PuO 2 ) fuel assemblies up to 50% of the core, together with UO 2 fuel assemblies. The use of MOX assemblies has a significant impact on key physic parameters and on safety. A lot of studies have been carried out in the past to explore the feasibility of plutonium recycling strategies by loading LWR reactors with MOX fuel. Many of these works were based on lattice codes, in order to perform detailed analyses of the neutronic characteristics of MOX assemblies. With the aim to take into account their interaction with surrounding UO 2 fuel elements, and the global effects on the core at operational conditions, an integrated approach making use of a 3-D core simulation is required. In this light, the present study adopts the state-of-art numerical models CASMO-5 and SIMULATE-3 to analyze the behavior of the core fueled with 30% MOX and to compare it with that of a large PWR reference core, fueled with UO 2 . (author)

  18. Preliminary studies leading to a conceptual design of a 1000 MWe fast neutron reactor

    International Nuclear Information System (INIS)

    Vendryes, G.; Zaleski, C.P.

    1964-01-01

    This report presents the results of studies which seemed important to undertake in connexion with the development of fast neutron reactors. - It points out the advantage of high internal breeding ratios (∼1, 1) which are necessary in order to get a small change in time both in power distribution and reactivity (less: than 0.005 Δk/k in 18 months). - It shows how to achieve this goal, when simultaneously power distribution flattening is obtained. These results in a higher mean specific power (which is an economic gain) and therefore in a smaller doubling time (about 10 years). - It attempts to find criteria concerning the specific power that should be used in future reactor designs -It presents a conceptional design of a 1000 MWe fast neutron reactor, for the realisation of which no technological impossibility appears. - It shows that the dynamic behaviour seems satisfactory despite a positive total isothermal sodium coefficient. - It tries to predict the development of fast reactors within the future total nuclear program. It does not appear that fissile materials supply problems should in France slow down the development of fast neutron reactors, which will be essentially tied up to its economical ability to produce cheap electric power. (authors) [fr

  19. IRSN-ANCCLI partnership. IRSN-ANCCLI seminar on decennial inspections of 900 MWe reactors - November 2010

    International Nuclear Information System (INIS)

    Rollinger, Francois; Delalonde, Jean-Claude; Hubert, F.; Paulmaz, X.; Tindillere, M.; Lheureux, Y.; Junker, R.; Sene, M.

    2010-11-01

    This seminar addressed the commitment of local information commissions (CLI) in the analysis and follow-up of the third decennial inspections of the French 900 MWe nuclear reactors. A first session addressed topics directly related to these inspections. Contributions proposed under the form of Power Point presentations by experts and representatives of the IRSN, EDF and CLIs addressed the following issues: safety re-examination of EDF 900 MWe reactors at the occasion of the third decennial inspection, activities of the IRSN related to skill management in nuclear power stations, implementation of the third decennial inspection of the unit 1 of the Fessenheim nuclear power station, the issue of follow-up by a local information commission after a decennial inspection. A second session addressed topics not related to decennial inspections and were proposed by Gravelines and Dampierre local information commissions: analysis of significant safety events, issues of skill management in nuclear power stations

  20. Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Choi, Seok Ki; Kim, Seong O

    2011-01-01

    A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

  1. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

    1982-01-01

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  2. Combustion and NOx emission characteristics of a retrofitted down-fired 660 MWe utility boiler at different loads

    Energy Technology Data Exchange (ETDEWEB)

    Li, Z.Q.; Liu, G.K.; Zhu, Q.Y.; Chen, Z.C.; Ren, F. [Harbin Institute of Technology, Harbin (China)

    2011-07-15

    Industrial experiments were performed for a retrofitted 660 MWe full-scale down-fired boiler. Measurements of ignition of the primary air/fuel mixture flow, the gas temperature distribution of the furnace and the gas components in the furnace were conducted at loads of 660, 550 and 330 MWe. With decreasing load, the gas temperature decreases and the ignition position of the primary coal/air flow becomes farther along the axis of the fuel-rich pipe in the burner region under the arches. The furnace temperature also decreases with decreasing load, as does the difference between the temperatures in the burning region and the lower position of the burnout region. With decreasing load, the exhaust gas temperature decreases from 129.8{sup o}C to 114.3{sup o}C, while NOx emissions decrease from 2448 to 1610 mg/m{sup 3}. All three loads result in low carbon content in fly ash and great boiler thermal efficiency higher than 92%. Compared with the case of 660 MWe before retrofit, the exhaust gas temperature decreased from 136 to 129.8{sup o}C, the carbon content in the fly ash decreased from 9.55% to 2.43% and the boiler efficiency increased from 84.54% to 93.66%.

  3. GTHTR 300 economic calculation with Mini G4ECONS as a basis for generation cost of GTHTR 10 MWe calculation

    International Nuclear Information System (INIS)

    Mochamad Nasrullah; Nurlaila

    2014-01-01

    The government plan to build Experimental Power Reactor (EPR) requires measurable economic assessment. The purpose of the study was to recalculate Gas Turbine High Temperature Reactor of 300 MWe (GTHTR 300) and compare the results with reference data. Then calculate generation cost of GTHTR 3, 5 and 10 MWe using the scale factor calculation. The methodology used is covered the generation cost calculation using the Mini G4Econs spread sheet models published by IAEA. Result of the verification calculation showed that a relatively similar, which means that the calculation model could be used to calculate for same other cases. Afterward, using scale factor, smaller scale reactor could be calculated. The calculation result show that electricity generation cost of SMR-HTR type with load factor 90% and discount rate 10% for power capacity 3, 5 and 10 MWe are 29.5, 22.68 and 16.17 cents$/kWh respectively. However, because the EPR is planning to be built as a non-commercial power reactors, then 5 % and 3 % of discount rate could be chosen, each of those discount rate will result electricity generation cost of 10.37 cents$/kWh and 8.56 cents$/kWh respectively. These result could be considered by the government for developing SMR type of HTR. (author)

  4. STUDY ON DISCHARGE HEAT UTILIZATION OF 250 MWe PCMSR TURBINE SYSTEM FOR DESALINATION USING MODIFIED MED

    Directory of Open Access Journals (Sweden)

    Andang Widiharto

    2015-03-01

    Full Text Available PCMSR (Passive Compact Molten Salt Reactor is one type of Advanced Nuclear Reactors. The PCMSR has benefit charasteristics of very efficient fuel use, high safety charecteristic as well as high thermodinamics efficiency. This is due to its breeding capability, inherently safe characteristic and totally passive safety system. The PCMSR design consists of three module, i.e. reactor module, turbine module and fuel management module. Analysis in performed by parametric calculation of the turbine system to calculate the turbine system efficiency and the hat available for desalination. After that the mass and energi balance of desalination process are calculated to calculate the amount of distillate produced and the amount of feed sea water needed. The turbine module is designed to be operated at maximum temperature cycle of 1373 K (1200 0C and minimum temperature cycle of 333 K (60 0K. The parametric calculation shows that the optimum turbine pressure ratio is 4.3 that gives the conversion efficiency of 56 % for 4 stages turbine and 4 stages compressor and equiped with recuperator. In this optimum condition, the 250 MWe PCMSR turbine system produces 196 MWth of waste heat with the temperature of cooling fluid in the range from 327 K (54 0C to 368 K (92 0C. This waste heat can be utilized for desalination. By using MMED desalination system, this waste heat can be used to produce fresh water (distillate from sea water feed. The amount of the destillate produced is 48663 ton per day by using 15 distillation effects. The performance ratio value is 2.8727 kg/MJ by using 15 distillation effects. Keywords: PCMSR, discharged heat, MMED desalination   PCMSR (Passive Compact Molten Salt Reactor merupakan salah satu tipe dari Reaktor Nuklir Maju. PCMSR memiliki keuntungan berupa penggunaan bahan bakar yang sangat efisisien, sifat keselamatan tinggi dan sekaligus efisiensi termodinamika yang tinggi. Hal ini disebabkan oleh kemampuan pembiakan bahan bakar, sifat

  5. PENGEMBANGAN MODEL UNTUK SIMULASI KESELAMATAN REAKTOR PWR 1000 MWe GENERASI III+ MENGGUNAKAN PROGRAM KOMPUTER RELAP5

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-04-01

    Full Text Available Reaktor daya PWR AP1000 yang didesain oleh Westinghouse adalah reaktor Generasi III+ pertama yang telah menerima persetujuan desain dari U.S. Nuclear Regulatory Commission (NRC. Saat ini utilitas China telah memulai pembangunan beberapa unit AP1000 di dua tapak terpilih untuk rencana operasi pada 2013-2015. AP1000 sebagai desain PWR berdasarkan teknologi teruji dari desain PWR lainnya yang dibuat oleh Westinghouse dengan penguatan pada sistem keselamatan pasif dengan demikian dapat dipertimbangkan untuk dibangun di Indonesia bila mengacu pada persyaratan pada PP 43/2006 mengenai Perijinan Reaktor Nuklir. Namun demikian, desain tersebut perlu diverifikasi oleh Technical Support Organization (TSO independen sebelum dapat dibangun di Indonesia. Verifikasi dapat dilakukan menggunakan paket program RELAP5 dalam bentuk analisis kecelakaan. Selama ini analisis kecelakaan PLTN dilakukan untuk tipe PWR 1000 MWe dari generasi II atau tipe konvensional. Mengingat saat ini referensi yang menggambarkan teknologi AP1000 yang menyertakan teknologi keselamatan pasif sudah tersedia maka dilakukan kegiatan pemodelan yang nantinya dapat digunakan untuk melakukan analisis kecelakaan. Metode pengembangan model mengacu pada pedoman IAEA yang terdiri dari pengumpulan data instalasi, pengembangan engineering data dan penyusunan input deck, verifikasi dan validasi data input. Model yang berhasil dikembangkan secara umum telah mewakili sistem AP1000 secara keseluruhan dan dianggap sebagai model dasar. Model tersebut telah diverifikasi dan divalidasi dengan data desain yang terdapat pada referensi dimana respon parameter termohidraulika menunjukkan perbedaan hasil ± 3% selain untuk parameter penurunan tekanan teras yang lebih rendah 13%. Sebagai model dasar, input deck yang diperoleh dapat dikembangkan lebih lanjut dengan mengintegrasikan pemodelan sistem keselamatan, sistem proteksi, dan sistem kendali yang spesifik AP1000 untuk keperluan simulasi keselamatan yang lebih

  6. Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2015-02-15

    Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  7. Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

    Science.gov (United States)

    Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

    2018-02-01

    Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

  8. Effect of flow configuration on moderator temperature distribution for 700 MWe Calandria

    International Nuclear Information System (INIS)

    Bharj, Jaspal Singh; Sahaya, R.R.; Dharne, S.P.

    2009-01-01

    The Calandria of a Pressurized Heavy Water Reactor (PHWR) is essentially a horizontal cylindrical vessel housing a matrix of horizontal tubes called Calandria tubes within which is contained the pressure tubes that house the fuel bundles. In addition there are horizontal and vertical flux control and shutdown devices. The Calandria is filled with heavy water moderator at a pressure slightly above the atmosphere. A large amount of heat (about 125 MWth) is generated within the moderator mainly due to neutron slowing down and attenuation of gamma radiations. This heat generation gives rise to a strong buoyancy-driven natural convection flow. In the proposed configuration of 700 MWe PHWR Calandria, moderator inlet diffusers are directed upwards and the outlet nozzles are at the bottom of the Calandria. The basis for the above said inlet/outlet configuration depends upon the various factors like space availability, NPSH requirement for the moderator pumps, and interference of flow with the other components inside the Calandria. This configuration is not conducive for the buoyancy-dominated flows generated due to large volumetric heat generation in the moderator. In order to see the effects of changes in flow configuration by re-orienting the inlet/outlet, a CFD study was undertaken for moderator flows in the conceptual Calandria. In the study, the moderator inlet diffusers direct the cool moderator towards the bottom of the Calandria and hot moderator flows out through the outlets in the upper half of the Calandria. The results of the study with various flow configurations show that modification in moderator flow configuration in Calandria, by way of introduction of moderator in the downward direction through diffusers and provision of the exits from the upper portion of the Calandria, results in significant reduction of the maximum temperature of moderator in Calandria. Further, the temperature distribution in the Calandria in the proposed configurations is much more

  9. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  10. Indian Summer

    Energy Technology Data Exchange (ETDEWEB)

    Galindo, E. [Sho-Ban High School, Fort Hall, ID (United States)

    1997-08-01

    This paper focuses on preserving and strengthening two resources culturally and socially important to the Shoshone-Bannock Indian Tribe on the Fort Hall Reservation in Idaho; their young people and the Pacific-Northwest Salmon. After learning that salmon were not returning in significant numbers to ancestral fishing waters at headwater spawning sites, tribal youth wanted to know why. As a result, the Indian Summer project was conceived to give Shoshone-Bannock High School students the opportunity to develop hands-on, workable solutions to improve future Indian fishing and help make the river healthy again. The project goals were to increase the number of fry introduced into the streams, teach the Shoshone-Bannock students how to use scientific methodologies, and get students, parents, community members, and Indian and non-Indian mentors excited about learning. The students chose an egg incubation experiment to help increase self-sustaining, natural production of steelhead trout, and formulated and carried out a three step plan to increase the hatch-rate of steelhead trout in Idaho waters. With the help of local companies, governmental agencies, scientists, and mentors students have been able to meet their project goals, and at the same time, have learned how to use scientific methods to solve real life problems, how to return what they have used to the water and land, and how to have fun and enjoy life while learning.

  11. Study of Epidemiology Conducted in Indian Nuclear Power Plants-Occupational Workers and Family Members

    International Nuclear Information System (INIS)

    Ramamirtham, B.; Shringi, K.; Wagh, P. M.

    2004-01-01

    At present in India, a nuclear generation capacity of 2720 MWe is in operation with 12 units of pressurised heavy water reactors (PHWRs) and 2 units of boiling water reactors (BWRs). The nature of the effects of the low-doses from ionizing radiation has been the subject considerable interest in the scientific community. The radiation exposures due to the operation of the NPPs are small and at low dose rates. The specific objective of the study were to compute the morbidity (prevalence) of cancer among the radiation occupational workers and their families and to compare with suitable controls and to study prevalence of congenital anomalies among the offspring of the employees of the nuclear power plants in India and to determine, if any, their causal relation with radiation exposure. The data collection work was carried out for survey by the local academic medical institutions near the NPP sites under the guidance of Tata Memorial Hospital, Mumbai. The distribution of the confounding factors among the radiation and non-radiation workers did not show any significant difference and thus the possibility of biased results was minimized. The cross-sectional survey has shown that there was no difference in the prevalence of malignancies in the radiation workers as compared to non-radiation workers, nor was there any difference in the prevalence of malignancies in the radiation workers as compared to non-radiation workers, nor was there any difference in the prevalence of malignancies in spouses and offspring. The study did not show any excess cancers among the study population. The congenital abnormalities observed in the offspring of the employees were much less than the reported values among the newborn children. The study has provided useful indicators and generated reliable baseline data for carrying out further work. Scientific thematic Area: 1) Radiation Effects. (Author)

  12. CONCEPTUAL DESIGN AND ECONOMICS OF A NOMINAL 500 MWe SECOND-GENERATION PFB COMBUSTION PLANT

    Energy Technology Data Exchange (ETDEWEB)

    A. Robertson; H. Goldstein; D. Horazak; R. Newby

    2003-09-01

    Research has been conducted under United States Department of Energy Contract DE-AC21-86MC21023 to develop a new type of coal-fired plant for electric power generation. This new type of plant, called a Second Generation Pressurized Fluidized Bed Combustion Plant (2nd Gen PFB), offers the promise of efficiencies greater than 48 percent, with both emissions and a cost of electricity that are significantly lower than those of conventional pulverized coal-fired (PC) plants with wet flue gas desulfurization. The 2nd Gen PFB plant incorporates the partial gasification of coal in a carbonizer, the combustion of carbonizer char in a pressurized circulating fluidized bed boiler, and the combustion of carbonizer syngas in a gas turbine combustor to achieve gas turbine inlet temperatures of 2300 F and higher. A conceptual design and an economic analysis was previously prepared for this plant. When operating with a Siemens Westinghouse W501F gas turbine, a 2400psig/1000 F/1000 F/2-1/2 in. Hg. steam turbine, and projected carbonizer, PCFB, and topping combustor performance data, the plant generated 496 MWe of power with an efficiency of 44.9 percent (coal higher heating value basis) and a cost of electricity 22 percent less than a comparable PC plant. The key components of this new type of plant have been successfully tested at the pilot plant stage and their performance has been found to be better than previously assumed. As a result, the referenced conceptual design has been updated herein to reflect more accurate performance predictions together with the use of the more advanced Siemens Westinghouse W501G gas turbine. The use of this advanced gas turbine, together with a conventional 2400 psig/1050 F/1050 F/2-1/2 in. Hg. steam turbine increases the plant efficiency to 48.2 percent and yields a total plant cost of $1,079/KW (January 2002 dollars). The cost of electricity is 40.7 mills/kWh, a value 12 percent less than a comparable PC plant.

  13. Improvements on computerized procedure system of advanced power reactor 1400 MWe

    International Nuclear Information System (INIS)

    Seong, Nokyu; Jung, Yeonsub; Sung, Chanho; Kang, Sungkon

    2017-01-01

    Plant procedures are instructions to help operator in monitoring, decision making, and controlling Nuclear Power Plants (NPPs). While plant procedures conventionally have been paper-based, computerized-based procedures are being implemented to reduce the drawbacks of paper-based procedures in many nuclear power plants. The Computerized Procedure System (CPS) designed by Korea Hydro and Nuclear Power Central Research Institute (KHNP CRI) is one of the human-system interfaces (HSIs) in digitalized Main Control Room (MCR) of APR1400 (Advanced Power Reactor 1400 MWe). Currently, CPS is being applied to constructing nuclear power plants of Korea and Barakah NPP 1, 2, 3 and 4 units of United Arab Emirates. The CPS has many advantages to perform the procedure in fully digitalized MCR. First, CPS provides the procedure flow with logic diagram to operators. The operator easily can be aware of the procedure flow from a previous instruction to the next instruction and also can find out the relation between parent instruction and child instructions such as AND, OR and SEQUENCE logics. Second, CPS has three logic-based functions such as procedure entry condition monitoring logic, continuously applied step (CAS) re-execution monitoring logic and auto evaluation logic on instructions. E.g. CPS provides the standard post trip actions procedure open popup message when the reactor trips by calculating the entry condition logic that procedure writer had made in the writing process. Third, CPS can directly display the task information related to instructions such as valves, pumps, process parameters, etc. and also the operator can call the system display related to procedure execution. If an operator clicks the system display link, the related system display popups on the right side monitor of CPS display. Lastly, CPS supports the synchronization of procedure among the operators. This synchronization function helps operators to succeed the goal of procedure and improve the situation

  14. Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-04-01

    The design, arrangement, thermodynamics, and economics of a 592 MW(e) (nominal gross) electric power generating plant equipped with a Babcock and Wilcox Company (B and W) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

  15. Ageing of fibre reinforced polymer composite selected as a bearing material for Rams of 540 MWe fuelling machine

    International Nuclear Information System (INIS)

    Limaye, P.K.; Soni, N.L.; Agrawal, R.G.

    2006-01-01

    Fibre-reinforced-polymer-composite material has been suggested as a bearing material to overcome tribological problems witnessed during the testing of Ram assembly of the 540 MWe fuelling machine at RTD. After successful trials at B-Ram the composite material has been adapted for B-RAM, C-Ram and RDB head at fuelling machines being tested at RTD, Hall 7 and at Tarapur. Laboratory evaluations were also carried out at Tribology Lab RTD to study effect of radiation on the composite. Paper deals with the various aspects of life prediction of this material in term of wear and radiation damage. (author)

  16. Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

    International Nuclear Information System (INIS)

    Mansouri, S.

    1985-06-01

    This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

  17. Technical notes for the conceptual design for an atmospheric fluidized-bed direct combustion power generating plant. [570 MWe plant

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-04-01

    The design, arrangement, thermodynamics, and economics of a 578 MW(e) (nominal gross) electric power generating plant equipped with a Foster Wheeler Energy Corporation (FWEC) atmospheric fluidized bed (AFB) boiler are described. Information is included on capital and operating costs, process systems, electrical systems, control and instrumentation, and environmental systems. This document represents a portion of an overall report describing the conceptual designs of two atmospheric fluidized bed boilers and balance of plants for the generation of electric power and the analysis and comparison of these conceptual designs to a conventional pulverized coal-fired electric power generation plant equipped with a wet limestone flue gas desulfurization system.

  18. Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

    International Nuclear Information System (INIS)

    Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

    1984-01-01

    The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

  19. Control rod studies for alternative fuel cycles in the GA 1160 MW(e) high temperature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Neef, H. J.

    1975-06-15

    The control system, which is investigated in this paper for both the low enriched uranium high enriched uranium/thorium fuel cycles, has been developed to control the General Atomics (GA) thorium fuel cycle 1160 MW(e) reactor. It has been shown in this investigation that its effectiveness in the low enriched and subsequent thorium cycle switch-over reactor is equivalent to the effectiveness in the thorium cycle. The shutdown margin in the low enriched core is even higher compared to the thorium core, mainly due to the presence of Pa-233 in the thorium cycle. As long as the fuel cycle for the thorium cycle is not closed with the recycling of U-233, the low enriched cycle will offer an attractive alternative. It was found that the GA 1160 MW(e) control system has enough built-in control rod capacity to accommodate the low enriched uranium cycle and to perform a later switch-over to a thorium-based fuel cycle.

  20. Evaluation of Two 300 MWe Fourth Generation Pb-Bi Reactor System Concepts

    International Nuclear Information System (INIS)

    Miller, Laurence F.; Khuram Khan, M.; Williams, Wesley; Mynatt, F.R.

    2002-01-01

    This paper describes the evaluation of two 300 MWe modular Pb-Bi cooled reactor system concepts that can be field assembled from components shipped on standard rail cars or on trucks. Thus, the largest components must be smaller than 12' x 12' x 80' (3.66 m x 3.66 m x 24.4 m) and should weigh no more than 80 tons. One of these systems utilizes a cylindrical two-loop containment vessel for the core and the other is a slab design. The fuel for both designs consists of standard-sized metallic IFR fuel in 17 x 17 square array assemblies with a pitch-to-diameter ratio of 1.15. The coolant outlet temperature is limited by current material technology, which is estimated to be 550 C. The primary coolant inlet temperature is selected to be 350 C. This is well above the melting temperature of Pb-Bi, and it is expected to be sufficiently high to limit transient-induced thermal stresses to acceptable values. Coolant flow rates through the core and external piping are below 1 m/s. The results from neutronics calculations include power distributions, reactivity coefficients, and fuel depletion, and results from heat transfer calculations include temperatures and flow rates at various locations in the primary and secondary systems. The neutronic design calculations are accomplished by using a discrete ordinate transport code and a cross section processing system developed at Oak Ridge National Laboratory. Two-dimensional flux distributions are obtained with the DOORS system, and ORIGEN-S, coupled with KENO, is used for time-dependent depletion calculations. The thermal-hydraulic design of the core consists of heat transfer and fluid flow calculation for an average channel. The inlet and outlet temperatures, along with the fuel centerline temperature, are determined in conjunction with core flow rates, pumping power, and total power output. This is accomplished by using a lumped parameter steady-state model with a spreadsheet and by using a one-dimensional time-dependent model

  1. 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

  2. Indian Ledger Art.

    Science.gov (United States)

    Chilcoat, George W.

    1990-01-01

    Offers an innovative way to teach mid-nineteenth century North American Indian history by having students create their own Indian Ledger art. Purposes of the project are: to understand the role played by American Indians, to reveal American Indian stereotypes, and to identify relationships between cultures and environments. Background and…

  3. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 4: Supplementary engineering data

    Science.gov (United States)

    1981-01-01

    The reference conceptual design of the Magnetohydrodynamic Engineering Test Facility (ETF), a prototype 200 MWe coal-fired electric generating plant designed to demonstrate the commercial feasibility of open cycle MHD is summarized. Main elements of the design are identified and explained, and the rationale behind them is reviewed. Major systems and plant facilities are listed and discussed. Construction cost and schedule estimates, and identification of engineering issues that should be reexamined are also given. The latest (1980-1981) information from the MHD technology program are integrated with the elements of a conventional steam power electric generating plant. Supplementary Engineering Data (Issues, Background, Performance Assurance Plan, Design Details, System Design Descriptions and Related Drawings) is presented.

  4. Magnetohydrodynamics MHD Engineering Test Facility ETF 200 MWe power plant. Conceptual Design Engineering Report CDER. Volume 3: Costs and schedules

    Science.gov (United States)

    1981-01-01

    The estimated plant capital cost for a coal fired 200 MWE electric generating plant with open cycle magnetohydrodynamics is divided into principal accounts based on Federal Energy Regulatory Commision account structure. Each principal account is defined and its estimated cost subdivided into identifiable and major equipment systems. The cost data sources for compiling the estimates, cost parameters, allotments, assumptions, and contingencies, are discussed. Uncertainties associated with developing the costs are quantified to show the confidence level acquired. Guidelines established in preparing the estimated costs are included. Based on an overall milestone schedule related to conventional power plant scheduling experience and starting procurement of MHD components during the preliminary design phase there is a 6 1/2-year construction period. The duration of the project from start to commercial operation is 79 months. The engineering phase of the project is 4 1/2 years; the construction duration following the start of the man power block is 37 months.

  5. Treatment and processing of the effluents and wastes (other than fuel) produced by a 900 MWe nuclear power plant

    International Nuclear Information System (INIS)

    Giraud

    1983-01-01

    Effluents produced by a 900 MWe power plant, are of three sorts: gaseous, liquid and solid. According to their nature, effluents are either released or stored for decaying before being released to the atmosphere. The non-contaminated reactor coolant effluents are purified (filtration, gas stripping) and treated by evaporation for reuse. Depending upon their radioactive level, liquid waste is either treated by evaporation or discharged after filtration. Solid waste issuing from previous treatments (concentrates, resins, filters) is processed in concrete drums using an encapsulation process. The concrete drum provides biological self-protection consistent with the national and international regulations pertaining to the transport of radioactive substance. Finally, the various low-level radioactive solid waste collected throughout the plant, is compacted into metal drums. Annual estimates of the quantity of effluents (gaseous, liquid) released in the environment and the number of drums (concrete, metal) produced by the plant figure in the conclusion

  6. The 900 MWe water pressurized reactor safety re-examination at the occasion of their third decennial inspection

    International Nuclear Information System (INIS)

    2009-01-01

    This document reports the safety re-examination actions performed on the French 900 MWe water pressurized reactors. This process includes three stages. The first one is an inventory of safety, design and operation requirements which are defined or specified in different texts: regulations, rules, criteria and specifications. This leads to compliance studies with respect to these documents and by in situ inspections, and then to corrective recommendations. After presenting this process, the report deals with specific safety studies which are related to external or internal aggressions (fire, explosions, flooding, climate, seism), to accidental situations (primary circuit cold overpressure, severe accidents, containment, level 1 and 2 safety probabilistic studies, passive failure of safeguard circuits, vapour generator tube failure, and so on), to design and sizing of civil engineering works and systems (radioactivity measurement system, safety injection system, recirculation function liability, liability of the irradiated fuel deactivation pool cooling system)

  7. Security of nuclear power in operation. Results from the first PWR 900 MWe stages of Electricity of France (EDF)

    Energy Technology Data Exchange (ETDEWEB)

    Capel, R; Chaubaron, J F [Electricite de France, 93 - Saint-Denis. Service de la Production Thermique

    1980-06-01

    The security and reliability objectives of the PWR 900 MWe stages are acquiring particular importance in the present energetic and nuclear context. This article presents the general framework wherein the superintendence and maintenance of plant equipment are situated. E.D.F. applies to all of its activities, the assurance of quality principles. The General Rules of Operating constitute the basic document. The Operating Technical Specifications specify the conditions for the correct operating and safety of the installations. The Organization of Quality handbook sets the rules to be obeyed in the management of all operations. Examples from Fessenhein and Bugey illustrate the subject and elucidate the practical dimension of security. Lastly, the lessons of experience are recalled.

  8. Some aspects of optimising the reactor core for a 600 MW(e) high temperature helium turbine power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Presser, W

    1972-04-24

    For the HHT 600 MW(e) power plant a core design with Triagonal blocks containing 24 channels with directly cooled fuel pins was considered. The design was found to require a low HM loading in the fuel zone to achieve favourable economic merits. For low HM densities a strong incentive exists to aim for burn-ups between 80 and 100 GWd/t. At the present an average discharge irradiation of 80 GWd/t was thought feasible and a reference design with a HM density of 0.6 g/cm {sup 3} in both core zones was chosen. The optimisation is not likely to be upset by local hot channel effects as a special investigation into the influence of safety margins found no changes in fuel cycle economics.

  9. Seismic analysis of two 1050 mm diameter heavy water upgrading towers for 235 MWe Kaiga Atomic Power Plant Site

    International Nuclear Information System (INIS)

    Soni, R.S.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.; Narwaria, Suresh; Vardarajan, T.G.; Sadhukhan, H.K.

    1992-01-01

    This report deals with the analysis carried out for the evaluation of earthquake induced stresses and deflections in two 1050 mm diameter heavy water upgrading towers for Kaiga Atomic Power Plant Site. The analysis of upgrading tower has been carried out for two mutually perpendicular horizontal excitations and one vertical excitation applied simultaneously. The upgrading towers have been analysed using beam model taking into account soil-structure interaction. Response spectrum analysis has been carried out using site spectra for 235 MWe Kaiga reactor site. The seismic analysis has been performed for both the towers with supporting structure along with concrete pedestals and raft foundation. The towers have been checked for its stability due to compressive stresses to avoid buckling so that the nearby safety related structures are not geopardised in the event of safe shutdown earthquake (SSE) loading. (author). 14 refs., 12 figs., 18 tabs

  10. Model, parameter and code of environmental dispersion of gaseous effluent under normal operation from nuclear power plant with 600 MWe

    International Nuclear Information System (INIS)

    Hu Erbang; Gao Zhanrong

    1998-06-01

    The model of environmental dispersion of gaseous effluence under normal operation from a nuclear power plant with 600 MWe is established to give a mathematical expression of annual mean atmospheric dispersion factor under mixing release condition based on quality assessment of radiological environment for 30 years of Chinese nuclear industry. In calculation, the impact from calm and other following factors have been taken into account: mixing layer, dry and wet deposition, radioactive decay and buildings. The doses caused from the following exposure pathways are also given by this model: external exposure from immersion cloud and ground deposition, internal exposure due to inhalation and ingestion. The code is named as ROULEA. It contains four modules, i.e. INPUT, ANRTRI, CHIQV and DOSE for calculating 4-dimension joint frequency, annual mean atmospheric dispersion factor and doses

  11. Modeling and simulations of a 30 MWe solar electric generating system using parabolic trough collectors in Turkey

    Energy Technology Data Exchange (ETDEWEB)

    Usta, Yasemin [Anyl Asansor Ltd (Turkey)], email: syusta@gmail.com; Baker, Derek [Middle East Technical University (Turkey)], email: dbaker@metu.edu.tr; Kaftanoglu, Bilgin [Atilim University (Turkey)], email: bilgink@atilim.edu.tr

    2011-07-01

    With the energy crisis and the increasing concerns about climate change, the interest in concentrating solar power (CSP) systems is growing in Turkey. The aim of this paper is to develop a model of a CSP system using a field of parabolic trough collectors and to assess the predicted performance of the system. A model was developed for a 30MWe solar generating system in Antalya, Turkey, using TRNSYS software, the solar thermal electric components library and information on an existing system in Kramer Junction, California, United States. Annual simulations were then performed for both systems in Antalya and California using weather data. It was found that the predictions were in good agreement with published data. In addition results showed that Antalya's system would generate 30% less than Kramer Junction's system on an annual basis. This paper provides useful information on modeling and simulation of CSP systems.

  12. Study on the muon spectra at the depth of 570 m.w.e. underground with 100t scintillation detector

    International Nuclear Information System (INIS)

    Enikeev, R.I.; Zatsepin, G.T.; Korol'kova, E.V.; Kudryavtsev, V.A.; Mal'gin, A.S.; Ryazhskaya, O.G.; Khal'chugov, F.F.

    1988-01-01

    The experiment was carried out with 100-ton scintillation detector placed in the salt mine at the depth of 570 m.w.e. Detector measured the spectrum of energy release of electromagnetic cascades generated by muons underground. Electromagnetic and nuclear cascades were separated by the number of neutrons contained in the cascades. The measured spectrum of energy releases agrees with π- and K-meson spectrum with γ π,K =1.75±0.08 for muon energies at sea level E μ 0 > 0.7 TeV. The experimental data transformed to the vertical muon spectrum at sea level are in good agreement with the results of other works. The primary cosmic ray spectrum and the characteristics of pA-interactions up to energies of ∼ 100 TeV have not a changes which would lead to the increase of the γ π,K value higher than 1.85

  13. Living probabilistic safety assessment of French 1300 MWe PWR nuclear power plant unit: methodology, results and teaching

    International Nuclear Information System (INIS)

    Dubreuil Chambardel, A.; Villemeur, A.; Berger, J.P.; Moroni, J.M.

    1991-02-01

    Launched in 1986 by Electricite de France, the Probabilistic Safety Assessment of a French 1300 MWe Pressurized Water Reactor (called PSA 1300) was completed in 1989. The first objective was to assess the annual core damage frequency by identifying all the accident scenarii likely to contribute significantly to this frequency. The second objective of the study was to provide an automated computerized tool (software) for updating the assessment - in order to take new data and knowledge into account - and for performing numerous sensitivity studies easily. Its scope and characteristics render this study unique. Indeed, it required an effort amounting to 50 engineer-years. The results and the first lessons are presented in this paper. The PSA 1300 teachings will be extensively used for the design and operation of existing or future French nuclear power reactors

  14. Influence of declivitous secondary air on combustion characteristics of a down-fired 300-MWe utility boiler

    Energy Technology Data Exchange (ETDEWEB)

    Zhengqi Li; Feng Ren; Zhichao Chen; Zhao Chen; Jingjie Wang [Harbin Institute of Technology, Harbin (China). School of Energy Science and Engineering

    2010-02-15

    Industrial experiments were performed with a 300-MWe full-scale down-fired boiler. New data is reported for (i) gas temperature distributions within the primary air and coal mixture flows, (ii) gas compositions, such as O{sub 2}, CO, CO{sub 2} and NOx, and (iii) gas temperatures within the near-wall region. The data complements previously-obtained data from the same utility boiler before being modified by declination of the F-tier secondary air. By directing secondary air under the arches, the region where the primary air and pulverized coal mixture is ignited is brought forward within the boiler. Gas temperatures rose in the fuel-burning zone and fell in the fuel-burnout zone. As a result the quantity of unburned carbon in fly ash and the gas temperature at the furnace outlet were both lowered. 20 refs., 7 figs., 2 tabs.

  15. Effect of combustion characteristics on wall radiative heat flux in a 100 MWe oxy-coal combustion plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, S.; Ryu, C. [Sungkyunkwan Univ., Suwon (Korea, Republic of). School of Mechanical Engineering; Chae, T.Y. [Sungkyunkwan Univ., Suwon (Korea, Republic of). School of Mechanical Engineering; Korea Institute of Industrial Technology, Cheonan (Korea, Republic of). Energy System R and D Group; Yang, W. [Korea Institute of Industrial Technology, Cheonan (Korea, Republic of). Energy System R and D Group; Kim, Y.; Lee, S.; Seo, S. [Korea Electric Power Research Institute (KEPRI), Daejeon (Korea, Republic of). Power Generation Lab.

    2013-07-01

    Oxy-coal combustion exhibits different reaction, flow and heat transfer characteristics from air-coal combustion due to different properties of oxidizer and flue gas composition. This study investigated the wall radiative heat flux (WRHF) of air- and oxy-coal combustion in a simple hexahedral furnace and in a 100 MWe single-wall-fired boiler using computational modeling. The hexahedral furnace had similar operation conditions with the boiler, but the coal combustion was ignored by prescribing the gas properties after complete combustion at the inlet. The concentrations of O{sub 2} in the oxidizers ranging between 26 and 30% and different flue gas recirculation (FGR) methods were considered in the furnace. In the hexahedral furnace, the oxy-coal case with 28% of O{sub 2} and wet FGR had a similar value of T{sub af} with the air-coal combustion case, but its WRHF was 12% higher. The mixed FGR case with about 27% O{sub 2} in the oxidizer exhibited the WRHF similar to the air-coal case. During the actual combustion in the 100 MWe boiler using mixed FGR, the reduced volumetric flow rates in the oxy-coal cases lowered the swirl strength of the burners. This stretched the flames and moved the high temperature region farther to the downstream. Due to this reason, the case with 30% O{sub 2} in the oxidizers achieved a WRHF close to that of air-coal combustion, although its adiabatic flame temperature (T{sub af}) and WHRF predicted in the simplified hexahedral furnace was 103 K and 10% higher, respectively. Therefore, the combustion characteristics and temperature distribution significantly influences the WRHF, which should be assessed to determine the ideal operating conditions of oxy- coal combustion. The choice of the weighted sum of gray gases model (WSGGM) was not critical in the large coal-fired boiler.

  16. Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

    International Nuclear Information System (INIS)

    Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

    2009-01-01

    The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

  17. Control rod cluster drop time anomaly Guandong nuclear power station (Daya bay) and Electricite de France nuclear power stations (1450 MWe N4 Series)

    International Nuclear Information System (INIS)

    Olivera, J.J.; Naury, S.; Tricot, N.; Tran Dai, P.; Gama, J.M.

    1996-01-01

    The anomaly of control rod cluster drop time revealed at Guandong Nuclear Power Station in Daya Bay and in the Chooz B1 pilot unit for the N4 series, led to the replacement of the M1 type control rod cluster guide tubes with 1300 MWe PWR type guide tubes, adapted to the geometry of the Guandong reactors and the 1450 MWe reactors of the N4 series. The comparison of the drop times obtained with the 1300 MWe type control rod cluster guide 1300 MWe type control rod cluster guide tubes gave satisfactory results. These met the safety criterion for N4 series control rod cluster drop times (2.15 under hot shutdown conditions). The drop time tests which will be carried out in middle of and at the end of cycle 1 of Chooz B1 should make it possible to finally validate the solution already successfully implemented at Guandong. However, this anomaly has revealed the limits of representativeness of the experimental test loops with regard to the real reactor configuration. In view of this, it has been deemed necessary to ask Electricite de France to pursue its analysis both on the understanding of the phenomena which led to this anomaly and on the limits of the representativeness of the experimental test loops. (authors)

  18. 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)

  19. Meeting the physics design challenges of modern LWRs being inducted into the Indian nuclear power programme

    International Nuclear Information System (INIS)

    Jagannathan, V.; Pal, Usha; Karthikeyan, R.; Raj, Devesh; Srivastava, Argala; Khan, Suhail Ahmad

    2007-01-01

    Indian nuclear power programme started with the two Boiling Water Reactors (BWR) of 210 MWe capacity at Tarapur. Two VVER-1000 MWe reactors which are Pressurized Water Reactors (PWR) of Russian design are being constructed at Kudankulam, Tamilnadu and are expected to be commissioned by end 2008. There may be also a possibility of inducting some western PWRs in future. These reactors belong to the category of light water reactors (LWR). The LWRs are compact and have complex physical characteristics distinctly different from those of the Pressurized Heavy Water Reactors (PHWR) which, currently form the mainstay of our indigenous nuclear power programme. The physics design and analysis capability for the modern LWRs (BWR, PWR and VVER) has been developed at Light Water Reactors Physics Section, BARC. This paper presents the current state of art in this key technology area to meet the physics design and operation challenges when LWRs would be inducted in a major way into the Indian nuclear power programme and commence operating in the coming decades. (author)

  20. 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

  1. Evolution of new X and Y positioning system for 540 MWe PHWR fuelling machines - based on commissioning experience

    International Nuclear Information System (INIS)

    Gupta, Vivek; Vyas, A.K.; Gupta, K.S.; Rama Mohan, N.; Bhambra, H.S.

    2006-01-01

    In PHWR units, X and Y positioning system is provided to give feedback regarding the misalignment between end-fitting and Fuelling Machine (FM) Head during homing on process for carrying out the correction before clamping the Head. The existing design of X and Y Positioning System works by measuring the misalignment by sensing the tilt of the FM Head in X and Y direction caused by its mechanical interfacing with end-fitting as it is advanced in Z direction. The misalignment of Head is corrected by moving it in X and Y direction by X-fine and Y-fine drives, at Z pre-stop position. This correction is vital for achieving the satisfactory sealing of heavy water from channel at snout of FM Head with end fitting. During testing and commissioning trials, it was found that the end fitting of 540 MWe coolant channel assembly either tilts or bends due to the application of load by Fuelling Machines during the process of homing-on of FM Head. Due to this phenomenon, value of misalignment sensed by the Positioning System was considerably lower than the actual misalignment and consequently results in uncorrected misalignment. It was also observed that the high unbalanced moments caused by movement of heavier mass of B-ram in FM Head was further aggravating the misalignment problem. The problem, as an interim measure, was solved by optimising the loads acting on the end fitting to achieve the practically minimum possible uncorrected misalignment. However, to provide a lasting solution for this problem, a new X and Y Positioning System has been evolved. In this system, the misalignment between FM Head and end fitting is found by direct actuation of linear Variable Differential Transformer (LVDT) sensors by four separate alignment plates mounted on the snout. Further development to evolve a completely non-invasive technique using laser sensors has also been undertaken. This paper describes the problems encountered during commissioning of existing design of X and Y Positioning

  2. Methods and results of a PSA level 2 for a German BWR of the 900 MWe class

    International Nuclear Information System (INIS)

    Loffler, H.; Sonnenkalb, M.

    2006-01-01

    On behalf of the federal Ministry for Environment, Nature Conservation and Reactor Safety (BMU) GRS has performed a PSA level 2 for a BWR type 69 NPP of the 900 MWe class, equipped with a N 2 inerted steel containment and a pressure suppression system. Integral deterministic accident analyses have been performed with the computer code MELCOR 1.8.5. Additional analyses have been done for those events and phenomena which are not or not sufficiently covered by MELCOR. The probabilistic event tree analysis begins with the core damage states received from PSA level 1, and it ends with the definition of release categories and the determination of their frequencies. Uncertainties about the frequency of core damage states and about events during the accident progression are taken into account by means of Monte Carlo simulations. If there is a core damage state there is a high probability (>50 %) for a very high and rapid release of radionuclides into the environment. This high conditional probability is due to the very low probability to retain a partly destroyed core inside the reactor pressure vessel (RPV) and because the containment almost certainly fails at the bottom of the control rod drives room after melt release from the failed RPV. (authors)

  3. Stress analysis of secondary ramp and secondary tilting mechanism of inclined fuel transfer machine for 500 MWe PFBR

    International Nuclear Information System (INIS)

    Prabhakaran, K.M.; Vaze, K.K.; Ghosh, A.K.; Rai, Somesh; Sundarani, A.R.; Patel, R.J.; Agrawal, R.G.

    2004-10-01

    Inclined Fuel Transfer Machine (IFTM) is one of the important machine of the fuel handling system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It is used to transfer core sub-assemblies (CSA) from reactor vessel to fuel building and vice-versa. Secondary ramp and Secondary tilting mechanism (SR/STM) is a part of IFTM which acts as a passage to transfer CSA. This mechanism and components were designed by the Refuelling Technology Division of BARC as per the ASME design code as class 2 component. Being critical in nature and complicated in geometry it was required to check the design of these components by detailed finite element analysis. The loading considered in the present study was static, thermal and seismic conditions. This was done using FEM software COSMOS/M. The Stresses were categorised as per the requirement of the ASME code for various levels of loading (Level A, B and C). Based on the analysis performed, it was concluded that the SR/STM qualifies the requirement of ASME code Section-III NC (Class-2 components). This report gives the details of the studies performed. (author)

  4. Effect of Flow Configuration on Velocity and Temperature Distribution of Moderator Inside 540 MWe PHWR Calandria using CFD Techniques

    International Nuclear Information System (INIS)

    Bharj, J.S.; Sahaya, R.R.; Datta, D.; Dharne, S.P.

    2006-01-01

    The calandria of a Pressurized Heavy Water Reactor (PHWR) is a horizontal cylindrical vessel housing a matrix of horizontal tubes called calandria tubes, through which pass the pressure tubes that house the fuel bundles. The calandria is filled with heavy water acting as moderator. A large amount of heat (about 95 MW) is generated within the moderator mainly due to neutron slowing down and attenuation of gamma radiations. In the present configuration of 540 MWe calandria, moderator inlet diffusers are directed upwards and the outlet is from the bottom of the calandria. This configuration is not conducive for the buoyancy-dominated flows generated due to large volumetric heat generation in the moderator. In order to decide the effects of changes in flow configuration by changing location/direction of inlet/outlet nozzles, a study was done for moderator flows in the using PHOENICS CFD software. The results of study with various flow configurations show that modification in moderator flow configuration, reduces the peak temperature of moderator in calandria by about 12 deg C as well as gives a much more uniform temperature distribution. (authors)

  5. Test of small-scale central-core-cavity closure for a 300-MW(e) GCFR

    International Nuclear Information System (INIS)

    Robinson, G.C.; Dougan, J.R.; Naus, D.J.

    1981-01-01

    Under the Prestressed Concrete Reactor Vessel (PCRV) Program at the Oak Ridge National Laboratory, model tests are conducted to verify the design of the PCRV for a 300 MW(e) Gas-Cooled Fast Reactor (GCFR). Prominent features of the 1:20-scale central core cavity model included a close pitched array of fifty-five penetration tubes, forty-four segmented gusset/bearing plate assemblies, and intermeshed reinforcing steel. The closure model which was designed for a maximum cavity pressure (MCP) of 10.08 MPa was initially tested by applying 10 pressurization cycles from essentially no load to the MCP with strain and deflection data obtained during each cycle. This was followed by pressurization cycles to 32.8 MPa, 41.3 MPa, 48.3 MPa, 58.4 MPa and 79.3 MPa. At a pressure of 79.3 MPa an end cap on a penetration tube developed leaks and the test was terminated. An inelastic analysis was conducted to provide an estimate of the ultimate strength of the closure plug and to determine the potential mode of failure

  6. Design of a multivariable controller for a CANDU 600 MWe nuclear power plant using the INA method

    International Nuclear Information System (INIS)

    Roy, N.; Boisvert, J.; Mensah, S.

    1984-04-01

    The development of large and complex nuclear and process plants requires high-performance control systems, designed with rigorous multivariable techniques. This work is part of an analytical study demonstrating the real potential of multivariable methods. It covers every step in the design of a multi-variable controller for a Gentilly-2 type CANDU 600 MWe nuclear power plant using the Inverse Nyquist Array (INA) method. First the linear design model and its preliminary modifications are described. The design tools are reviewed and the operations required to achieve open-loop diagonal dominance are thoroughly described. Analysis of the closed-loop system is then performed and a feedback matrix is selected to meet the design specifications. The performance of the controller on the linear model is verified by simulation. Finally, the controller is implemented on the reference non-linear model to assess its overall performance. The results show that the INA method can be used successfully to design controllers for large and complex systems

  7. Mathematical modelling of straw combustion in a 38 MWe power plant furnace and effect of operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yao Bin Yang; Robert Newman; Vida Sharifi; Jim Swithenbank; John Ariss [Sheffield University, Sheffield (United Kingdom). Sheffield University Waste Incineration Centre (SUWIC), Department of Chemical and Process Engineering

    2007-01-15

    As one of the most easily accessible renewable energy resources, straw can be burned to provide electricity and heat to local communities. In this paper, mathematical modelling methods have been employed to simulate the operation of a 38 MWe straw-burning power plant to obtain detailed information on the flow and combustion characteristics in the furnace and to predict the effect on plant performance of variation in operating conditions. The predicted data are compared to measurements in terms of burning time, furnace temperature, flue gas emissions (including NOx), carbon content in the ash and overall combustion efficiency. It is concluded that straw burning on the grate is locally sub-stoichiometric and most of the NO is formed in the downstream combustion chamber and radiation shaft; auxiliary gas burners are responsible for the uneven distribution of temperature and gas flow at the furnace exit; and fuel moisture content is limited to below 25% to prevent excessive CO emission without compromising the plant performance. The current work greatly helps to understand the operating characteristics of large-scale straw-burning plants. 33 refs., 15 figs., 3 tabs.

  8. A through calculation of 1,100 MWe PWR large break LOCA by THYDE-P1 EM model

    International Nuclear Information System (INIS)

    Kanazawa, Masayuki; Asahi, Yoshiro; Hirano, Masashi

    1984-07-01

    THYDE-P1 is a code to analyze both the blowdown and refill-reflood phases of loss-of-coolant accidents (LOCAs) of pressurized water reactors (PWRs). Up to now, THYDE-P1 has been applied to various experiment analyses, which show its high capability to analyze LOCAs as a best estimate (BE) calculation code. In this report, evaluation model (EM) calculation method, especialy in the blowdown and refill phases, is established equivalently to WREM/J2 which is regarded as appropriate for an EM calculation code, and the results of them are compared and discussed. The present calculation was the first executed by THYDE-P1-EM, and was performed as Sample Calculation Run 80 which was a part of a series of THYDE-P sample calculations. The calculation was carried out from the LOCA initiation till 400 seconds for a guillotine break at the cold leg of a commercial 1,100 MWe PWR plant. The calculated results agreed well to that of the WREM/J2 code. (author)

  9. Structural mechanics research and development for main components of chinese 300 MWe PWR NPPs: from design to life management

    International Nuclear Information System (INIS)

    Yao Weida; Dou Yikang; Xie Yongcheng; He Yinbiao; Zhang Ming; Liang Xingyun

    2005-01-01

    Qinshan Nuclear Power Plant (Unit I), is a 300 MWe prototype PWR independently developed by Chinese own efforts, from design, manufacture, construction, installation, commissioning, to operation, inspection, maintenance, ageing management and lifetime assessment. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) has taken up with and involved in deeply the R and D to tackle problems of this type of reactor since very beginning in early 1970s. Structural mechanics is one of the important aspects to ensure the safety and reliability for NPP components. This paper makes a summary on role of structural mechanics for component safety and reliability assessment in different stages of design, commissioning, operation, as well as lifetime assessment on this type PWR NPPs, including Qinshan-I and Chashma-I, a sister plant in Pakistan designed by SNERDI. The main contents of the paper cover design by analysis for key components of NSSS; mechanical problems relating to safety analysis; special problems relating to pressure retaining components, such as fracture mechanics, sealing analysis and its test verifications, etc.; experimental research on flow-induced vibration; seismic qualification for components; component failure diagnosis and root cause analysis; vibration qualification and diagnosis technique; component online monitoring technique; development of defect assessment; methodology of aging management and lifetime assessment for key components of NPPs, etc. (authors)

  10. The Influence of atmospheric conditions to probabilistic calculation of impact of radiology accident on PWR 1000 MWe

    International Nuclear Information System (INIS)

    Pande Made Udiyani; Sri Kuntjoro

    2015-01-01

    The calculation of the radiological impact of the fission products releases due to potential accidents that may occur in the PWR (Pressurized Water Reactor) is required in a probabilistic. The atmospheric conditions greatly contribute to the dispersion of radionuclides in the environment, so that in this study will be analyzed the influence of atmospheric conditions on probabilistic calculation of the reactor accidents consequences. The objective of this study is to conduct an analysis of the influence of atmospheric conditions based on meteorological input data models on the radiological consequences of PWR 1000 MWe accidents. Simulations using PC-Cosyma code with probabilistic calculations mode, the meteorological data input executed cyclic and stratified, the meteorological input data are executed in the cyclic and stratified, and simulated in Muria Peninsula and Serang Coastal. Meteorological data were taken every hour for the duration of the year. The result showed that the cumulative frequency for the same input models for Serang coastal is higher than the Muria Peninsula. For the same site, cumulative frequency on cyclic input models is higher than stratified models. The cyclic models provide flexibility in determining the level of accuracy of calculations and do not require reference data compared to stratified models. The use of cyclic and stratified models involving large amounts of data and calculation repetition will improve the accuracy of statistical calculation values. (author)

  11. Design of 50 MWe HTR-PBMR reactor core and nuclear power plant fuel using SRAC2006 programme

    International Nuclear Information System (INIS)

    Bima Caraka Putra; Yosaphat Sumardi; Yohannes Sardjono

    2014-01-01

    This research aims to assess the design of core and fuel of nuclear power plant type High Temperature Reactor-Pebble Bed Modular Reactor 50 MWe from the Beginning of Life (BOL) to Ending of life (EOL) with eight years operating life. The parameters that need to be analyzed in this research are the temperature distribution inside the core, quantity enrichment of U 235 , fuel composition, criticality, and temperature reactivity coefficient of the core. The research was conducted with a data set of core design parameters such as nuclides density, core and fuel dimensions, and the axial temperature distribution inside the core. Using SRAC2006 program package, the effective multiplication factor (k eff ) values obtained from the input data that has been prepared. The results show the value of the criticality of core is proportional to the addition of U 235 enrichment. The optimum enrichment obtained at 10.125% without the use of burnable poison with an excess reactivity of 3.1 2% at BOL. The addition Gd 2O3 obtained an optimum value of 12 ppm burnable poison with an excess reactivity 0.38 %. The use of Er 2O3 with an optimum value 290 ppm has an excess reactivity 1.24 % at BOL. The core temperature reactivity coefficient with and without the use of burnable poison has a negative values that indicates the nature of its inherent safety. (author)

  12. Development of high pressure conductivity probe (HPCP) for secondary shut down system (SDS-2) of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Mohan, L.R.

    2003-09-01

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface. This interface moves towards the calandria because of molecular diffusion, temperature difference and physical disturbances in the moderator level. It is proposed to install two numbers of high pressure conductivity probes (HPCP) to monitor the interface movement as well as to provide the safe annunciation value for interface location. On actuation of the SDS-2 signal, high-pressure helium will inject the poison into the moderator to shutdown the reactor. During poison injection, these probes will experience high pressure of nearly 85 kg/sq.cm. Global market survey indicated that conductivity probes having built in temperature sensor are available for a maximum pressure rating of 35 kg/sq.cm. Hence in order to meet the process requirement of SDS-2, the development of HPCP suitable for a pressure of 85 kg/sq.cm. was taken up. Two numbers of such probes were successfully designed, fabricated and evaluated for their performance. The developed conductivity probes fully meet the laid design and performance criteria. The aforesaid development work was a successful endeavour towards indigenisation of high-pressure conductivity probe for future applications. This report deals with the design aspects, fabrication technique, material and performance evajuation criteria and test results of HPCP. (author)

  13. Foster Wheeler's Solutions for Large Scale CFB Boiler Technology: Features and Operational Performance of Łagisza 460 MWe CFB Boiler

    Science.gov (United States)

    Hotta, Arto

    During recent years, once-through supercritical (OTSC) CFB technology has been developed, enabling the CFB technology to proceed to medium-scale (500 MWe) utility projects such as Łagisza Power Plant in Poland owned by Poludniowy Koncern Energetyczny SA. (PKE), with net efficiency nearly 44%. Łagisza power plant is currently under commissioning and has reached full load operation in March 2009. The initial operation shows very good performance and confirms, that the CFB process has no problems with the scaling up to this size. Also the once-through steam cycle utilizing Siemens' vertical tube Benson technology has performed as predicted in the CFB process. Foster Wheeler has developed the CFB design further up to 800 MWe with net efficiency of ≥45%.

  14. The effect of core design changes on the doubling time and the fuel cycle cost of a 1,000 MWe LMFBR

    International Nuclear Information System (INIS)

    Otake, I.; Inoue, T.; Tomabechi, K.; Osada, H.; Aoki, K.

    1978-01-01

    Core design studies were performed to improve the doubling time and to minimize the fuel cycle cost of a 1,000 MWe Fast Demonstration Reactor. A core was designed mainly based on the technology being used for the design of a prototype fast reactor MONJU, because much valuable experience will be forthcoming from this reactor. Design parameters with a wide variable range were used to clarify the relations between breeding characteristics, fuel economics and various designs. (author)

  15. Severe accident mitigation strategy for the generation II PWRs in France. Some outcomes of the on-going periodic safety review of the French 1300 MWe PWR series

    Energy Technology Data Exchange (ETDEWEB)

    Cenerino, G.; Rahni, N.; Chevrier, P.; Dubreuil, M.; Guigueno, Y.; Raimond, E.; Bonnet, J.M. [IRSN/PSN-RES/SAG (France)

    2013-07-01

    The 3{sup rd} Periodic Safety Review of the French 1300 MWe PWRs series includes some modifications to increase their robustness in case of a severe accident. Their review is based on both deterministic and probabilistic approaches, keeping in mind that severe accidents frequencies and radiological consequences should be as low as reasonably practicable, severe accidents management strategies should be as safe as possible and the robustness of equipment used for severe accident management should be ensured. Consequently, the IRSN level 2 probabilistic safety assessment (L2 PSA) studies for the 1300 MWe reactors have been used to re-assess the results of the utility's L2 PSA and rank them to identify the containment failure modes contributing the most to the global risk. This ranking helped the review of plant modifications. Regarding strategies for accident management, the EDF management of water in the reactor cavity during a severe accident for the 1300 MWe PWRs is presented as well as the IRSN position on this strategy: this is an example where the optimal severe accident management strategy choice is not so easy to define. Regarding the robustness of equipment used for severe accident management, the interest of a diversification or redundancy of the French emergency filtered containment venting opening is one example among many others. (orig.)

  16. Leadership Preferences of Indian and Non-Indian Athletes.

    Science.gov (United States)

    Malloy, D. C.; Nilson, R. N.

    1991-01-01

    Among 86 Indian and non-Indian volleyball competitors, non-Indian players indicated significantly greater preferences for leadership that involved democratic behavior, autocratic behavior, or social support. Indians may adapt their behavior by participating in non-Indian games, without changing their traditional value orientations. Contains 22…

  17. 75 FR 61511 - Indian Gaming

    Science.gov (United States)

    2010-10-05

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs.... FOR FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office of the.... SUPPLEMENTARY INFORMATION: Under section 11 of the Indian Gaming Regulatory Act of 1988 (IGRA), Public Law 100...

  18. 76 FR 42722 - Indian Gaming

    Science.gov (United States)

    2011-07-19

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs... Date: July 19, 2011. FOR FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming... INFORMATION: Under section 11 of the Indian Gaming Regulatory Act of 1988 (IGRA), Public Law 100-497, 25 U.S.C...

  19. 75 FR 38834 - Indian Gaming

    Science.gov (United States)

    2010-07-06

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs...: July 6, 2010. FOR FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office...-4066. SUPPLEMENTARY INFORMATION: Under Section 11 of the Indian Gaming Regulatory Act of 1988 (IGRA...

  20. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Editorial Board. Sadhana. Editor. N Viswanadham, Indian Institute of Science, Bengaluru. Senior Associate Editors. Arakeri J H, Indian Institute of Science, Bengaluru Hari K V S, Indian Institute of Science, Bengaluru Mujumdar P P, Indian Institute of Science, Bengaluru Manoj Kumar Tiwari, Indian Institute of Technology, ...

  1. Primary system thermal hydraulics of future Indian fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K., E-mail: kvelu@igcar.gov.in [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Natesan, K.; Maity, Ram Kumar; Asokkumar, M.; Baskar, R. Arul; Rajendrakumar, M.; Sarathy, U. Partha; Selvaraj, P.; Chellapandi, P. [Thermal Hydraulics Section, Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 (India); Kumar, G. Senthil; Jebaraj, C. [AU-FRG Centre for CAD/CAM, Anna University, Chennai 600 025 (India)

    2015-12-01

    Highlights: • We present innovative design options proposed for future Indian fast reactor. • These options have been validated by extensive CFD simulations. • Hotspot factors in fuel subassembly are predicted by parallel CFD simulations. • Significant safety improvement in the thermal hydraulic design is quantified. - Abstract: As a follow-up to PFBR (Indian prototype fast breeder reactor), many FBRs of 500 MWe capacity are planned. The focus of these future FBRs is improved economy and enhanced safety. They are envisaged to have a twin-unit concept. Design and construction experiences gained from PFBR project have provided motivation to achieve an optimized design for future FBRs with significant design changes for many critical components. Some of the design changes include, (i) provision of four primary pipes per primary sodium pump, (ii) inner vessel with single torus lower part, (iii) dome shape roof slab supported on reactor vault, (iv) machined thick plate rotating plugs, (v) reduced main vessel diameter with narrow-gap cooling baffles and (vi) safety vessel integrated with reactor vault. This paper covers thermal hydraulic design validation of the chosen options with respect to hot and cold pool thermal hydraulics, flow requirement for main vessel cooling, inner vessel temperature distribution, safety analysis of primary pipe rupture event, adequacy of decay heat removal capacity by natural convection cooling, cold pool transient thermal loads and thermal management of top shield and reactor vault.

  2. Evaluation on the habitability of a reactor control room for a 1300 MWe PWR following a LOCA

    International Nuclear Information System (INIS)

    Chang, Si Young; Ha, Chung Woo

    1988-01-01

    An evaluation on the habitability of a reactor control room for a French 1300 MWe P'4 type PWR following a LOCA has been performed through exposure dose assessment for a reactor operator. A computer code COREX calculating the time-integrated exposure dose has been developed to provide a reasonable basis in this evaluation. Using COREX the exposure dose reduction factors in the reactor control room, the time--integrated radioactivities released into the atmosphere and the time-integrated exposure dose up to 30 days following the LOCA can be also calculated. From the exposure dose assessment, the time-integrated exposure dose to whole body and thyroid of a reactor operator were 0.36 mSv(0.036 rem) and 480 mSv(48.0 rem), respectively after 30 days following the LOCA. The thyroid dose of 480 mSv was nearly 10 times greater than the dose equivalent limit of 50 mSv(5.0 rem) set by the ICRP. Regarding the habitability of a reactor control room, this exceeding thyroid exposure dose could be reduced to 1.2 mSv(0.12 rem), which is 400 times less than the original, by considering the practical 4 work-shifts a day, and by improving the iodine removal efficiency of the filtration system n the reactor control room through the reinforcement of charcoal bed filters for iodine removal. The radiological habitability of a reactor control room, therefore, could be assured by comparing with the dose equivalent limit of the ICRP

  3. Process simulation of co-firing torrefied biomass in a 220 MWe coal-fired power plant

    International Nuclear Information System (INIS)

    Li, Jun; Zhang, Xiaolei; Pawlak-Kruczek, Halina; Yang, Weihong; Kruczek, Pawel; Blasiak, Wlodzimierz

    2014-01-01

    Highlights: • The performances of torrefaction based co-firing power plant are simulated by using Aspen Plus. • Mass loss properties and released gaseous components have been studied during biomass torrefaction processes. • Mole fractions of CO 2 and CO account for 69–91% and 4–27% in total torrefied gases. • The electrical efficiency reduced when increasing either torrefaction temperature or substitution ratio of biomass. - Abstract: Torrefaction based co-firing in a pulverized coal boiler has been proposed for large percentage of biomass co-firing. A 220 MWe pulverized coal-power plant is simulated using Aspen Plus for full understanding the impacts of an additional torrefaction unit on the efficiency of the whole power plant, the studied process includes biomass drying, biomass torrefaction, mill systems, biomass/coal devolatilization and combustion, heat exchanges and power generation. Palm kernel shells (PKS) were torrefied at same residence time but 4 different temperatures, to prepare 4 torrefied biomasses with different degrees of torrefaction. During biomass torrefaction processes, the mass loss properties and released gaseous components have been studied. In addition, process simulations at varying torrefaction degrees and biomass co-firing ratios have been carried out to understand the properties of CO 2 emission and electricity efficiency in the studied torrefaction based co-firing power plant. According to the experimental results, the mole fractions of CO 2 and CO account for 69–91% and 4–27% in torrefied gases. The predicted results also showed that the electrical efficiency reduced when increasing either torrefaction temperature or substitution ratio of biomass. A deep torrefaction may not be recommended, because the power saved from biomass grinding is less than the heat consumed by the extra torrefaction process, depending on the heat sources

  4. Thermoeconomic Modeling and Parametric Study of a Photovoltaic-Assisted 1 MWe Combined Cooling, Heating, and Power System

    Directory of Open Access Journals (Sweden)

    Alexandros Arsalis

    2016-08-01

    Full Text Available In this study a small-scale, completely autonomous combined cooling, heating, and power (CCHP system is coupled to a photovoltaic (PV subsystem, to investigate the possibility of reducing fuel consumption. The CCHP system generates electrical energy with the use of a simple gas turbine cycle, with a rated nominal power output of 1 MWe. The nominal power output of the PV subsystem is examined in a parametric study, ranging from 0 to 600 kWe, to investigate which configuration results in a minimum lifecycle cost (LCC for a system lifetime of 20 years of service. The load profile considered is applied for a complex of households in Nicosia, Cyprus. The solar data for the PV subsystem are taken on an hourly basis for a whole year. The results suggest that apart from economic benefits, the proposed system also results in high efficiency and reduced CO2 emissions. The parametric study shows that the optimum PV capacity is 300 kWe. The minimum lifecycle cost for the PV-assisted CCHP system is found to be 3.509 million €, as compared to 3.577 million € for a system without a PV subsystem. The total cost for the PV subsystem is 547,445 €, while the total cost for operating the system (fuel is 731,814 € (compared to 952,201 € for a CCHP system without PVs. Overall the proposed system generates a total energy output of 210,520 kWh (during its whole lifetime, which translates to a unit cost of 17 €/kWh.

  5. Burnup effects on criticality, breeding and safety of 1,000 MWe gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Ohta, Fumio

    1977-12-01

    Burnup characteristics of 1,000 MWe, PuO 2 - UO 2 fuelled helium-cooled fast breeder reactor have been studied concerning criticality, breeding and safety. A 26-energy group cross-section set produced from ENDF/B-3 was used. Criticality and breeding were studied with two-dimensional burnup code APOLLO and 4-energy group cross-section set generated by collapsing the mentioned cross-section set. Safety aspects such as Doppler reactivity effect, coolant-depressurisation and steam-ingression reactivity effect were studied with multi-dimensional diffusion theory code CITATION and perturbation theory code PERKY, as well as the 26-energy group cross-section set. The following were revealed: (1) The reactivity swing over a year's irradiation is merely 1.5% ΔK/K. This small swing may permit relatively long fuel dwelling in GCFR and , thus, the frequency of outages for refuelling can be minimised. (2) The surplus fissile plutonium over a year's irradiation is about 360 Kg, and the system doubling time is about 9 years. The GCFR studied has excellent breeding, compared with those in PuO 2 -UO 2 fuelled LMFBR and other GCFRs. (3) The coolant-depressurisation reactivity effect becomes more positive with burnup. This is not so serious as the sodium-void reactivity effect of LMFBR. (4) In the start-up core, the steam-ingression reactivity effect due to steam ingression to the core and blanket from the secondary coolant system becomes positive at certain steam density (0.02gr/cc) and this positive effect increases with steam density. With advance of burnup, however, the effect becomes negative, this increasing with steam density. After all, the steam ingression is no hazard in operation of GCFR since the reactivity effect is negative in the equilibrium state. (auth.)

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

    International Nuclear Information System (INIS)

    Zheng Yanhua; Shi Lei; Wang Yan

    2010-01-01

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

  7. Thermal hydraulic conditions inducing incipient cracking in the 900 MWe unit 93 D reactor coolant pump shafts

    International Nuclear Information System (INIS)

    Bore, C.

    1995-01-01

    From 1987, 900 MWe plant operating feedback revealed cracking in the lower part of the reactor coolant pump shafts, beneath the thermal ring. Metallurgical examinations established that this was due to a thermal fatigue phenomenon known as thermal crazing, occurring after a large number of cycles. Analysis of thermal hydraulic conditions initiating the cracks does not allow exact quantification of the thermal load inducing cracking. Only qualitative analyses are thus possible, the first of which, undertaken by the pump manufacturer, Jeumont Industrie, showed that the cracks could not be due to the major transients (stop-start, injection cut-off), which were too few in number. Another explanation was then put forward: the thermal ring, shrunk onto the shaft it is required to protect against thermal shocks, loosens to allow an alternating downflow of cold water from the shaft seals and an upflow of hot water from the primary system. However, approximate calculations showed that the flow involved would be too slight to initiate the cracking observed. A more stringent analysis undertaken with the 2D flow analysis code MELODIE subsequently refuted the possibility of alternating flows beneath the ring establishing that only a hot water upflow occurred due to a 'viscosity pump' phenomenon. Crack initiation was finally considered to be due to flowrate variations beneath the ring, with the associated temperature fluctuations. This flowrate fluctuation could be due to an unidentified transient phenomenon or to a variation in pump operating conditions. This analysis of the hydraulic conditions initiating the cracks disregards shaft surface residual stresses. These are tensile stresses and show that loads less penalizing than those initially retained could cause incipient cracking. Thermal ring modifications to reduce these risks were proposed and implemented. In addition, final metallurgical treatment of the shafts was altered and implemented. In addition, final metallurgical

  8. Experimental study of poison moderator interface movement for shut down system #2(SDS#2) of 540 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.; Chawan, D.B.; Ananthan, P.; Sharma, B.S.V.G.; Mohan, L.R.

    2005-03-01

    The poison solution and the moderator in Secondary Shutdown System (SDS-2) of 500 MWe PHWR, are separated by their own liquid in liquid interface, termed as poison moderator interface (PMI). During normal operation of the reactor, the interface moves towards the calandria, mainly because of molecular diffusion from poison to moderator. Other reasons for movement are mixing of poison and moderator due to physical disturbances in the moderator level and to some extent due to temperature difference between the two liquids. The electrical conductivity of these liquids was found to be the most reliable parameter indicating interface movement. For this purpose, two on-line high-pressure conductivity probes have been installed on moderator side for each one of the six poison tanks. During normal operation of reactor, the interface moves slowly towards the calandria over a period of time and gives rise to increase in conductivity. To study the interface pattern and factors affecting the same, a full-scale experimental setup was developed and series of experiments carried out. The experimental results showed that the interface is quite stable and annunciation can be placed around 100 micro siemens/cm before back flushing is initiated. One dimensional diffusion analysis of the obtained experimental data showed that the derived model for PMI setup with diffusion parameter of 900 cm 2 /hr is able to predict the interface movement quite satisfactorily. This report gives an insight into the experiments carried out for estimation of the effective diffusion parameter for the poison moderator interface, model formulation and its prognostic behavior. (author)

  9. Engineered safety in development of liquid poison injection system (shut down system-2) for 500 MWe PHWR

    International Nuclear Information System (INIS)

    Sapra, M.K.; Kundu, S.N.; Mohan, L.R.

    2002-01-01

    Full text: The provision of shut down systems (SDS) is a mandatory requirement for safety of any nuclear reactor. The SDS shall be capable of making and holding the core adequately subcritical in the event of any anticipated operational occurrence and postulated accident conditions. The shut down function will perform as intended when its design and components are thoroughly evaluated for their reliability and effectiveness. A full scale mock up for one injection unit was designed and developed at Hall No.7, BARC. Experimental studies were carried out to qualify the design and evolve process parameters such as gas tank pressure, poison discharge rate and poison injection time. In liquid poison injection system i.e. shutdown system -2, there is no physical barrier, between the two liquids i.e. the poison and the moderator. A liquid in liquid interface, called poison moderator interface (PMI) separates these fluids. Extensive lab scale studies have been carried out on PMI movement study i.e. the interface movement due to molecular diffusion and due to process disturbances under simulated reactor condition. On the basis of lab scale results, a full-scale PMI setup has been designed and developed to generate plant data. From reactor safety consideration, the floating ball in poison tank is designed in such a way that it prevents the over pressurisation of calandria. For this purpose a non-intrusive ultrasonic ball detection system (U-BDS) has been developed. This paper covers the PMI system for 500 MWe PHWR with relevant safety aspects and describes in detail, the experimental results of PMI study. The engineered safety in design, methodology and qualification of U-BDS and its role intended in performance of SDS-2 have been also discussed in the paper

  10. Theoretical and experimental investigations into natural circulation behaviour in a simulated facility of the Indian PHWR under reduced inventory conditions

    International Nuclear Information System (INIS)

    Satish Kumar, N.V.; Nayak, A.K.; Vijayan, P.K.; Pal, A.K.; Saha, D.; Sinha, R.K.

    2004-01-01

    A theoretical and experimental investigation has been carried out to study natural circulation characteristics of an Indian PHWR under reduced inventory conditions. The theoretical model incorporates a quasi-steady state analysis of natural circulation at different system inventories. It predicts the system flow rate under single-phase and two-phase conditions and the inventory at which reflux condensation occurs. The model predictions were compared with test data obtained from FISBE (facility for integral system behaviour experiments), which simulates the thermal hydraulic behaviour of the Indian 220 MWe PHWR. The experimental results were found to be in close agreement with the predictions. It was also found that the natural circulation could be oscillatory under reduced inventory conditions. (orig.)

  11. About | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    The 82nd Annual Meeting of the Indian Academy of Sciences is being held at ... by newly elected Fellows and Associates over a wide range of scientific topics. ... Indian Institute of Science Education and Research (IISER), Bhopal: Indian ...

  12. Indianization of psychiatry utilizing Indian mental concepts

    Science.gov (United States)

    Avasthi, Ajit; Kate, Natasha; Grover, Sandeep

    2013-01-01

    Most of the psychiatry practice in India is guided by the western concepts of mental health and illness, which have largely ignored the role of religion, family, eastern philosophy, and medicine in understanding and managing the psychiatric disorders. India comprises of diverse cultures, languages, ethnicities, and religious affiliations. However, besides these diversities, there are certain commonalities, which include Hinduism as a religion which is spread across the country, the traditional family system, ancient Indian system of medicine and emphasis on use of traditional methods like Yoga and Meditation for controlling mind. This article discusses as to how mind and mental health are understood from the point of view of Hinduism, Indian traditions and Indian systems of medicine. Further, the article focuses on as to how these Indian concepts can be incorporated in the practice of contemporary psychiatry. PMID:23858244

  13. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Dept. of Electrical Engineering, Indian Institute of Technology, Kandi, ... Specialization: Elementary Particle Physics Address during Associateship: Centre for Theoretical Studies, Indian Institute of Science, Bangalore 560 012.

  14. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Director, Indian Institute of Science Education & Research, .... Address: Visiting Professor, CORAL, Indian Institute of Technology, ..... Specialization: Elementary Particles & High Energy Physics, Plasma Physics and Atomic Physics

  15. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Department of Chemistry, Indian Institute of Technology, Powai, Mumbai .... Address: Emeritus Professor, National Institute of Advanced Studies, Indian .... Specialization: High Energy & Elementary Particle Physics, Supersymmetric ...

  16. Evolution of on-power refuelling system for 500 MWe PHWR based on experience from Rajasthan, Madras and Narora Atomic Power Stations

    International Nuclear Information System (INIS)

    Warrier, S.R.; Inder Jit; Sanatkumar, A.

    1991-01-01

    The on-power fuel handling system design at Rajasthan and Madras Atomic Power Stations (RAPS and MAPS) is essentially based on the design of the fuel handling system at Douglas Point Station (CANADA) Although, a number of improvements have been carried out in the fuel handling system of RAPS and MAPS at the component and sub-assembly level, some problems of repetitive nature like frequent deterioration in the performance of B-ram ball screw, leak detector solenoid valves etc., still exist. Further, there are certain limitations and drawbacks in the fuelling systems of these stations. For example, FM carriage design would not meet current seismic qualification standards. Also there are chances of fuel transfer room getting contaminated during movement of a failed fuel bundle. In order to obviate these deficiencies, a new concept has been worked out for the fuel handling system of Narora Atomic Power Station (NAPS) and accordingly, major changes have been made adopting a new layout. For example, FM head supporting arrangement has been changed to 'Suspension' type and a 'Linear-indexed' transfer magazine has been introduced in the fuel transfer system. Based on the experience gained from RAPS, MAPS and NAPS, design concept for 500 MWe fuel handling system has been evolved with further improvements especially in the layout. Also, a Calibration and Maintenance Facility for maintenance, testing calibration of FM head, sub-assemblies and components of fuel handling system has been introduced in the 500 MWe design. This paper discusses some of the experience gained from RAPS, MAPS and NAPS and also highlights the features of 500 MWe fuel handling system. (author)

  17. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 2: Engineering. Volume 3: Costs and schedules

    Science.gov (United States)

    1981-01-01

    Engineering design details for the principal systems, system operating modes, site facilities, and structures of an engineering test facility (ETF) of a 200 MWE power plant are presented. The ETF resembles a coal-fired steam power plant in many ways. It is analogous to a conventional plant which has had the coal combustor replaced with the MHD power train. Most of the ETF components are conventional. They can, however, be sized or configured differently or perform additional functions from those in a conventional coal power plant. The boiler not only generates steam, but also performs the functions of heating the MHD oxidant, recovering seed, and controlling emissions.

  18. Magnetohydrodynamics (MHD) Engineering Test Facility (ETF) 200 MWe power plant. Conceptual Design Engineering Report (CDER). Volume 2: Engineering. Volume 3: Costs and schedules. Final Report

    International Nuclear Information System (INIS)

    1981-09-01

    Engineering design details for the principal systems, system operating modes, site facilities, and structures of an engineering test facility (ETF) of a 200 MWE power plant are presented. The ETF resembles a coal-fired steam power plant in many ways. It is analogous to a conventional plant which has had the coal combustor replaced with the MHD power train. Most of the ETF components are conventional. They can, however, be sized or configured differently or perform additional functions from those in a conventional coal power plant. The boiler not only generates steam, but also performs the functions of heating the MHD oxidant, recovering seed, and controlling emissions

  19. Stopped cosmic-ray muons in plastic scintillators on the surface and at the depth of 25 m.w.e

    International Nuclear Information System (INIS)

    Maletić, D; Dragić, A; Banjanac, R; Joković, D; Veselinović, N; Udovicić, V; Savić, M; Anicin, I; Puzović, J

    2013-01-01

    Cosmic ray muons stopped in 5 cm thick plastic scintillators at surface and at depth of 25 m.w.e are studied. Apart from the stopped muon rate we measured the spectrum of muon decay electrons and the degree of polarization of stopped muons. Preliminary results for the Michel parameter yield values lower than the currently accepted one, while the asymmetry between the numbers of decay positrons registered in the upper and lower hemispheres appear higher than expected on the basis of numerous earlier studies.

  20. Red Women, White Policy: American Indian Women and Indian Education.

    Science.gov (United States)

    Warner, Linda Sue

    This paper discusses American Indian educational policies and implications for educational leadership by Indian women. The paper begins with an overview of federal Indian educational policies from 1802 to the 1970s. As the tribes have moved toward self-determination in recent years, a growing number of American Indian women have assumed leadership…

  1. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast ...

  2. In-calandria retention of corium in Indian PHWR - experimental simulations with decay heat

    International Nuclear Information System (INIS)

    Nayak, A.K.

    2015-01-01

    The severe accident at Fukushima has compelled the nuclear community to relook at the safety of existing nuclear power plants (NPP) against natural origin events of beyond design basis and prolonged station black out (SBO). A major lesson learned is to assess the capability of the safety systems to cool the reactor core and spent fuel storage facilities in the event of a prolonged station black out (SBO). Similar safety review is planned for the Indian Pressurized Heavy Water Reactors (PHWRs) considering a prolonged SBO. The Indian PHWR is a heavy water-moderated and cooled, natural uranium-fuelled reactor in which the horizontal fuel channels are submerged in a pool of heavy water moderator located inside the calandria vessel. The calandria vessel is surrounded by a calandria vault having large volume of light water. Concerns are raised that in the event of an unmitigated SBO, it may result into a low probable severe accident leading to core melt down. The core melt may further fail the calandria vessel in case the melt is not quenched. If the calandria vessel fails, the corium shall interact with the cold calandria vault water and concrete resulting in generation of large amount of non-condensable gases and steam which will lead to over pressurization of containment and may cause its failure. Therefore, in-calandria corium retention via external cooling using vault water can be considered as an important accident management program in PHWR. In this strategy, the core melt retains inside the calandria vessel by continually removing the stored heat and decay heat through outer surface of the vessel by cooling water and maintaining the integrity of the vessel. The present study focuses on experimental investigation in a scaled facility of an Indian PHWR to investigate the coolability of molten corium with simulated decay heat by using the calandria vault water. Molten borosilicate glass was used as the simulant due to its comparable heat transfer characteristics

  3. Defeathering the Indian.

    Science.gov (United States)

    LaRoque, Emma

    In an effort to mitigate the stultified image of the American Indian in Canada, this handbook on Native Studies is written from the Indian point of view and is designed to sensitize the dominant society, particularly educators. While numerous approaches and pointers are presented and specific mateirals are recommended, the focus is essentially…

  4. American Indian Community Colleges.

    Science.gov (United States)

    One Feather, Gerald

    With the emergence of reservation based community colleges (th Navajo Community College and the Dakota Community Colleges), the American Indian people, as decision makers in these institutions, are providing Indians with the technical skills and cultural knowledge necessary for self-determination. Confronted with limited numbers of accredited…

  5. Indian Summer Arts Festival


    OpenAIRE

    Martel, Yann; Tabu; Tejpal, Tarun; Kunzru, Hari

    2011-01-01

    The SFU Woodward's Cultural Unit partnered with the Indian Summer Festival Society to kick off the inaugural Indian Summer Festival. Held at the Goldcorp Centre for the Arts, it included an interactive Literature Series with notable authors from both India and Canada, including special guests Yann Martel, Bollywood superstar Tabu, journalist Tarun Tejpal, writer Hari Kunzru, and many others.

  6. Indian Ocean Rim Cooperation

    DEFF Research Database (Denmark)

    Wippel, Steffen

    Since the mid-1990s, the Indian Ocean has been experiencing increasing economic cooperation among its rim states. Middle Eastern countries, too, participate in the work of the Indian Ocean Rim Association, which received new impetus in the course of the current decade. Notably Oman is a very active...

  7. The Indian Monsoon

    Indian Academy of Sciences (India)

    Pacific Oceans, on subseasonal scales of a few days and on an interannual scale. ... over the Indian monsoon zone2 (Figure 3) during the summer monsoon .... each 500 km ×500 km grid over the equatorial Indian Ocean, Bay of Bengal and ...

  8. Indian Arts in Canada

    Science.gov (United States)

    Tawow, 1974

    1974-01-01

    A recent publication, "Indian Arts in Canada", examines some of the forces, both past and present, which are not only affecting American Indian artists today, but which will also profoundly influence their future. The review presents a few of the illustrations used in the book, along with the Introduction and the Foreword. (KM)

  9. 76 FR 49505 - Indian Gaming

    Science.gov (United States)

    2011-08-10

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Tribal-State Class III Gaming Compact taking effect. SUMMARY: This publishes..., Director, Office of Indian Gaming, Office of the Deputy Assistant Secretary--Policy and Economic...

  10. 75 FR 38833 - Indian Gaming

    Science.gov (United States)

    2010-07-06

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes... Date: July 6, 2010. FOR FURTHER INFORMATION CONTACT: Paula Hart, Director, Office of Indian Gaming...

  11. 77 FR 76513 - Indian Gaming

    Science.gov (United States)

    2012-12-28

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Amended Tribal-State Class III Gaming Compact taking effect. SUMMARY..., 2012. FOR FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office of the...

  12. 76 FR 165 - Indian Gaming

    Science.gov (United States)

    2011-01-03

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs... Wisconsin Gaming Compact of 1992, as Amended in 1999, 2000, and 2003. DATES: Effective Date: January 3, 2011. FOR FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office of the...

  13. 75 FR 68618 - Indian Gaming

    Science.gov (United States)

    2010-11-08

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs... of Wisconsin Gaming Compact of 1991, as Amended in 1999 and 2003. DATES: Effective Date: November 8, 2010. FOR FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office of the...

  14. 77 FR 76514 - Indian Gaming

    Science.gov (United States)

    2012-12-28

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact taking effect. SUMMARY: This... FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office of the Deputy...

  15. New associates | Announcements | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Sushmee Badhulika, Indian Institute of Technology, Hyderabad ... Sankar Chakma, Indian Institute of Science Education & Research, Bhopal Joydeep ... B Praveen Kumar, Indian National Centre for Ocean Information Services, Hyderabad

  16. Estimated radiological effects of the normal discharge of radioactivity from nuclear power plants in the Netherlands with a total capacity of 3500 MWe

    International Nuclear Information System (INIS)

    Lugt, G. van der; Wijker, H.; Kema, N.V.

    1977-01-01

    In the Netherlands discussions are going on about the installation of three nuclear power plants, leading with the two existing plants to a total capacity of 3500 MWe. To have an impression of the radiological impact of this program, calculations were carried out concerning the population doses due to the discharge of radioactivity from the plants during normal operation. The discharge via the ventilation stack gives doses due to noble gases, halogens and particulate material. The population dose due to the halogens in the grass-milk-man chain is estimated using the real distribution of grass-land around the reactor sites. It could be concluded that the population dose due to the contamination of crops and fruit is negligeable. A conservative estimation is made for the dose due to the discharge of tritium. The population dose due to the discharge in the cooling water is calculated using the following pathways: drinking water; consumption of fish; consumption of meat from animals fed with fish products. The individual doses caused by the normal discharge of a 1000 MWe plant appeared to be very low, mostly below 1 mrem/year. The population dose is in the order of some tens manrems. The total dose of the 5 nuclear power plants to the dutch population is not more than 70 manrem. Using a linear dose-effect relationship the health effects to the population are estimated and compared with the normal frequency

  17. Reactor cooling systems thermal-hydraulic assessment of the ASTEC V1.3 code in support of the French IRSN PSA-2 on the 1300 MWe PWRs

    International Nuclear Information System (INIS)

    Tregoures, Nicolas; Philippot, Marc; Foucher, Laurent; Guillard, Gaetan; Fleurot, Joelle

    2010-01-01

    The French Institut de Radioprotection et de Surete Nucleaire (IRSN) is performing a level 2 Probabilistic Safety Assessment (PSA-2) on the French 1300 MWe PWRs. This PSA-2 study is relying on the ASTEC integral computer code, jointly developed by IRSN and GRS (Germany). In order to assess the reliability and the quality of physical results of the ASTEC V1.3 code as well as the PWR 1300 MWe reference input deck, a wide-ranging series of comparisons with the French best-estimate thermal-hydraulic code CATHARE 2 V2.5 has been performed on 14 different severe-accident scenarios. The present paper details 4 out of the 14 studied scenarios: a 12-in. cold leg Loss of Coolant Accident (LOCA), a 2-tube Steam Generator Tube Rupture (SGTR), a 12-in. Steam Line Break (SLB) and a total Loss of Feed Water scenario (LFW). The thermal-hydraulic behavior of the primary and secondary circuits is thoroughly investigated and compared to the CATAHRE 2 V2.5 results. The ASTEC results of the core degradation phase are also presented. Overall, the thermal-hydraulic behavior given by the ASTEC V1.3 is in very good agreement with the CATHARE 2 V2.5 results.

  18. Rasam Indian Restaurant: Menu

    OpenAIRE

    Rasam Indian Restaurant

    2013-01-01

    Rasam Indian Restaurant is located in the Glasthule, a suburb of Dublin and opened in 2003. The objective is to serve high quality, authentic Indian cuisine. "We blend, roast and grind our own spices daily to provide a flavour that is unique to Rasam. Cooking Indian food is founded upon long held family traditions. The secret is in the varying elements of heat and spices, the tandoor clay oven is a hugely important fixture in our kitchen. Marinated meats are lowered into the oven on long m...

  19. [Indian workers in Oman].

    Science.gov (United States)

    Longuenesse, E

    1985-01-01

    Until recently Oman was a country of emigration, but by 1980 an estimated 200,000 foreign workers were in the country due to the petroleum boom. Almost 1/3 of the estimated 300,000 Indian workers in the Gulf states were in Oman, a country whose colonial heritage was closely tied to that of India and many of whose inhabitants still speak Urdu. The number of work permits granted to Indians working in the private sector in Oman increased from 47,928 in 1976 to 80,787 in 1980. An estimated 110,000 Indians were working in Oman in 1982, the great majority in the construction and public works sector. A few hundred Indian women were employed by the government of Oman, as domestics, or in other capacities. No accurate data is available on the qualifications of Indian workers in Oman, but a 1979 survey suggested a relatively low illiteracy rate among them. 60-75% of Indians in Oman are from the state of Kerala, followed by workers from the Punjab and the southern states of Tamil Nadu and Andhra Pradesh and Bombay. Indian workers are recruited by specialized agencies or by friends or relatives already employed in Oman. Employers in Oman prefer to recruit through agencies because the preselection process minimizes hiring of workers unqualified for their posts. Officially, expenses of transportation, visas, and other needs are shared by the worker and the employer, but the demand for jobs is so strong that the workers are obliged to pay commissions which amount to considerable sums for stable and well paying jobs. Wages in Oman are however 2 to 5 times the level in India. Numerous abuses have been reported in recruitment practices and in failure of employers in Oman to pay the promised wages, but Indian workers have little recourse. At the same level of qualifications, Indians are paid less then non-Omani Arabs, who in turn receive less than Oman nationals. Indians who remain in Oman long enough nevertheless are able to support families at home and to accumulate considerable

  20. Indian concepts on sexuality.

    Science.gov (United States)

    Chakraborty, Kaustav; Thakurata, Rajarshi Guha

    2013-01-01

    India is a vast country depicting wide social, cultural and sexual variations. Indian concept of sexuality has evolved over time and has been immensely influenced by various rulers and religions. Indian sexuality is manifested in our attire, behavior, recreation, literature, sculptures, scriptures, religion and sports. It has influenced the way we perceive our health, disease and device remedies for the same. In modern era, with rapid globalization the unique Indian sexuality is getting diffused. The time has come to rediscover ourselves in terms of sexuality to attain individual freedom and to reinvest our energy to social issues related to sexuality.

  1. Regression analysis of radiological parameters in nuclear power plants

    International Nuclear Information System (INIS)

    Bhargava, Pradeep; Verma, R.K.; Joshi, M.L.

    2003-01-01

    Indian Pressurized Heavy Water Reactors (PHWRs) have now attained maturity in their operations. Indian PHWR operation started in the year 1972. At present there are 12 operating PHWRs collectively producing nearly 2400 MWe. Sufficient radiological data are available for analysis to draw inferences which may be utilised for better understanding of radiological parameters influencing the collective internal dose. Tritium is the main contributor to the occupational internal dose originating in PHWRs. An attempt has been made to establish the relationship between radiological parameters, which may be useful to draw inferences about the internal dose. Regression analysis have been done to find out the relationship, if it exist, among the following variables: A. Specific tritium activity of heavy water (Moderator and PHT) and tritium concentration in air at various work locations. B. Internal collective occupational dose and tritium release to environment through air route. C. Specific tritium activity of heavy water (Moderator and PHT) and collective internal occupational dose. For this purpose multivariate regression analysis has been carried out. D. Tritium concentration in air at various work location and tritium release to environment through air route. For this purpose multivariate regression analysis has been carried out. This analysis reveals that collective internal dose has got very good correlation with the tritium activity release to the environment through air route. Whereas no correlation has been found between specific tritium activity in the heavy water systems and collective internal occupational dose. The good correlation has been found in case D and F test reveals that it is not by chance. (author)

  2. Indian refining industry

    International Nuclear Information System (INIS)

    Singh, I.J.

    2002-01-01

    The author discusses the history of the Indian refining industry and ongoing developments under the headings: the present state; refinery configuration; Indian capabilities for refinery projects; and reforms in the refining industry. Tables lists India's petroleum refineries giving location and capacity; new refinery projects together with location and capacity; and expansion projects of Indian petroleum refineries. The Indian refinery industry has undergone substantial expansion as well as technological changes over the past years. There has been progressive technology upgrading, energy efficiency, better environmental control and improved capacity utilisation. Major reform processes have been set in motion by the government of India: converting the refining industry from a centrally controlled public sector dominated industry to a delicensed regime in a competitive market economy with the introduction of a liberal exploration policy; dismantling the administered price mechanism; and a 25 year hydrocarbon vision. (UK)

  3. CFD analysis of the pulverized coal combustion processes in a 160 MWe tangentially-fired-boiler of a thermal power plant

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Cristiano V. da; Beskow, Arthur B. [Universidade Regional Integrada do Alto Uruguai e das Misses (LABSIM/GEAPI/URI), Erechim, RS (Brazil). Dept. de Engenharia e Ciencia da Computacao. Grupo de Engenharia Aplicada a Processos Industriais], Emails: cristiano@uricer.edu.br, Arthur@uricer.edu.br; Indrusiak, Maria Luiza S. [Universidade do Vale do Rio dos Sinos (UNISINOS), Sao Leopoldo, RS (Brazil). Programa de Engenharia Mecanica], E-mail: sperbindrusiak@via-rs.net

    2010-10-15

    The strategic role of energy and the current concern with greenhouse effects, energetic and exegetic efficiency of fossil fuel combustion greatly enhance the importance of the studies of complex physical and chemical processes occurring inside boilers of thermal power plants. The state of the art in computational fluid dynamics and the availability of commercial codes encourage numeric studies of the combustion processes. In the present work the commercial software CFX Ansys Europe Ltd. was used to study the combustion of coal in a 160 MWe commercial thermal power plant with the objective of simulating the operational conditions and identifying factors of inefficiency. The behavior of the flow of air and pulverized coal through the burners was analyzed, and the three-dimensional flue gas flow through the combustion chamber and heat exchangers was reproduced in the numeric simulation. (author)

  4. Electronuclear reactors - EDF - Orientations of generic studies to be performed for the safety re-examination of 1300 MWe reactors associated to their third decennial inspection

    International Nuclear Information System (INIS)

    2011-01-01

    This report expresses the ASN's opinion on the framework and objectives of the EDF program concerning the generic studies of the safety re-examination of the 1300 MWe reactors associated with their third decennial inspection. This comprises lessons from the Fukushima accident, the improvement of the 'internal explosion' referential by using a probabilistic study, the application of the seismic margin assessment approach as soon as possible, checking the absence of any 'cliff effect' for cooling functions, a deepened re-examination of hurricane frequencies in France. Other request by the ASN concern the verification of the pertinence of release authorizations, taking the TSN law into account, taking the AP1300 project into account, the expansion of the complementary domain, the project of reactor lifetime extension. Some technical requests are discussed in appendix

  5. Analysis for the coolability of the reactor cavity in a Korean 1000 MWe PWR using MELCOR 1.8.3 computer code

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Ju Yeul; Chung, Chang Hyun; Park, Soo Yong

    1996-01-01

    The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction (MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass. The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment

  6. Running-in strategies for the low-enriched 600 MW(e) D-HHT reactor. Part 1. Comparison of different on-load refuelling schemes

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U

    1973-03-14

    This paper presents detailed burn-up calculations and fuel management strategies for the Dragon-HHT, D-HHT, reference core. The reference layout was chosen from the outcome of a design survey with the 1-D equilibrium fuel cycle code FLATTER. The decision was based on aspects of engineering and economics. The purpose of the investigation is to devise a suitable first core, follow the irradiation history of the fuel and the general behaviour of the reactor during the first core replacements until equilibrium operating conditions are reached. A detailed description of time dependant burn-up and spatial power production for specified reactivity limits is required. For this purpose the reactor code system VSOP was employed. Different combinations of the parameters are investigated and the influence on reactor operation and economics discussed. From the strategy analysis a reference fuel management scheme is chosen for the low enriched 600 MW(e) D-HHT reactor.

  7. Opinion of the IRSN on serviceability of the 900 MWe reactor vessel - Answers to demands of the Nuclear Permanent Department of December 2005 - Mechanical aspect

    International Nuclear Information System (INIS)

    2010-05-01

    As three demands had been made to EDF in December 2005 regarding the serviceability and more particularly the mechanical behaviour of the 900 MWe reactor vessels, this report discusses the evolution brought to models and proposed by EDF to correct the defect plasticity and take residual stresses into account. This discussion notably concerns the defect height and length range, and the admissible residual stress, but also the use of safety coefficients, transient application, fluence and the brittle-ductile transition temperature. This report from the French Nuclear Safety and Radioprotection Institute (IRSN) outlines the failure risks associated to the vessel in some specific nuclear power stations. Recommendations are made regarding the residual stress amplitude, the risk of fracture by cleavage, and actions to correct fracture risk margins on vessels which do not comply with regulatory criteria

  8. Experimental study of the tritium inventory in the BR3 and extrapolation to a P.W.R. of 900 MWe

    International Nuclear Information System (INIS)

    Charlier, A.; Gubel, P.; Vandenberg, C.; Haas, D.

    1982-01-01

    The aim of this report is to evaluate the tritium production and diffusion in uranium and plutonium fuel in the primary circuit of a PWR and to improve the knowledge about the production difference between the two kinds of isotopes. The first part of the work is relative to the experimental PWR BR3, cycle 4A, during which a constant control of the tritium activity has been performed in the primary circuit. These experimental evaluation was compared with the the theoretical estimation of the tritium production during the cycle 4A. From these observations and calculations, a tritium release fraction was deduced and estimated to be 0.81% of the total tritium produced in the fuel. The second part of the work is devoted to post-irradiation examinations on a few uranium and plutonium rods irradiated in the BR3 reactor. The tritium content was measured in the cladding, in the fuel and in the gas plenum for various samples of fuel rods. These results show the relationship between the release rate from the fuel matrix, the linear power and the burnup. The last part of the work is the estimate of the tritium production in a PWR of 900 MWe in operating conditions. The tritium production was calculated for an uranium fuelled core and for a core containing 30% of all plutonium fuel assemblies in a generic power plant of 900 MWe. From this study, it results that the loading with 30% plutonium assemblies at equilibrium increases the tritium balance in the moderator water of less than 5%

  9. Safety and Performance Achievement of Indian Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kumar, Randhir

    2011-01-01

    The Nuclear Power Programme in India is based on three stage. The first stage is based on setting up of Pressurized Heavy Water Reactors (PHWRs) using indigenously available natural uranium producing electricity and plutonium. This will be followed by the second stage by plutonium fuelled Fast Breeder Reactors (FBRs) producing electricity and additional quantity of plutonium and also uranium 233 from thorium. The third stage of reactors will be based on thorium uranium 233 cycle.

  10. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Author Affiliations. A Salih1 S Ghosh Moulic2. Department of Aerospace Engineering, Indian Institute of Space Science and Technology, Thiruvananthapuram 695 022; Department of Mechanical Engineering, Indian Institute of Technology, Kharagpur 721 302 ...

  11. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Sequential Bayesian technique: An alternative approach for software reliability estimation ... Software reliability; Bayesian sequential estimation; Kalman filter. ... Department of Mathematics, Indian Institute of Technology, Kharagpur 721 302; Reliability Engineering Centre, Indian Institute of Technology, Kharagpur 721 302 ...

  12. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Director, Indian Institute of Science Education & Research, Sri Rama ... Address: Department of Chemistry, Indian Institute of Technology, New Delhi 110 016, Delhi ..... Specialization: Elementary Particle Physics, Field Theory and ...

  13. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Author Affiliations. Soumen Bag1 Gaurav Harit2. Department of Computer Science and Engineering, Indian Institute of Technology Kharagpur, Kharagpur 721 302, India; Information and Communication Technology, Indian Institute of Technology Rajasthan, Jodhpur 342 011, India ...

  14. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.2. Three-dimensional analysis of the temperature and stress fields in a HHT vessel, including effects of the thermal creep

    International Nuclear Information System (INIS)

    Rodriguez, C.; Rebora, B.

    1979-01-01

    The thermal rheological calculation of the prestressed concrete reactor vessel for the HHT-670 MW(e) Demonstration Plant is presented in the paper. The main aim of this calculation is to evaluate the effects of the elevated temperature and various loads on the liner as well as on the hot concrete

  15. 77 FR 5566 - Indian Gaming

    Science.gov (United States)

    2012-02-03

    ... up to 900 gaming devices, any banking or percentage card games, and any devices or games authorized... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Tribal--State Class III Gaming Compact Taking Effect. SUMMARY: This publishes...

  16. 76 FR 56466 - Indian Gaming

    Science.gov (United States)

    2011-09-13

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an approval of the gaming compact between the Flandreau Santee Sioux Tribe and the State of South...

  17. 76 FR 65208 - Indian Gaming

    Science.gov (United States)

    2011-10-20

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an Approval of the Gaming Compact between the Confederated Tribes of the [[Page 65209

  18. 75 FR 68823 - Indian Gaming

    Science.gov (United States)

    2010-11-09

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Amendment. SUMMARY: This notice publishes approval of the Amendments to the Class III Gaming Compact (Amendment) between the State of Oregon...

  19. 77 FR 43110 - Indian Gaming

    Science.gov (United States)

    2012-07-23

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an extension of Gaming between the Rosebud Sioux Tribe and the State of South Dakota. DATES...

  20. 75 FR 8108 - Indian Gaming

    Science.gov (United States)

    2010-02-23

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes... Governing Class III Gaming. DATES: Effective Date: February 23, 2010. FOR FURTHER INFORMATION CONTACT: Paula...

  1. 76 FR 8375 - Indian Gaming

    Science.gov (United States)

    2011-02-14

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes an extension of the Gaming Compact between the Oglala Sioux Tribe and the State of South Dakota...

  2. 78 FR 10203 - Indian Gaming

    Science.gov (United States)

    2013-02-13

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal State Class III Gaming Compact. SUMMARY: This notice publishes the Approval of the Class III Tribal- State Gaming Compact between the Chippewa-Cree Tribe of the...

  3. 77 FR 30550 - Indian Gaming

    Science.gov (United States)

    2012-05-23

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes approval by the Department of an extension to the Class III Gaming Compact between the Pyramid Lake Paiute...

  4. 77 FR 45371 - Indian Gaming

    Science.gov (United States)

    2012-07-31

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an extension of Gaming between the Oglala Sioux Tribe and the State of South Dakota. DATES: Effective...

  5. 76 FR 11258 - Indian Gaming

    Science.gov (United States)

    2011-03-01

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Tribal--State Class III Gaming Compact taking effect. SUMMARY: Notice is given that the Tribal-State Compact for Regulation of Class III Gaming between the Confederated Tribes of the...

  6. 78 FR 15738 - Indian Gaming

    Science.gov (United States)

    2013-03-12

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an extension of the gaming compact between the Rosebud Sioux Tribe and the State of South Dakota...

  7. 77 FR 41200 - Indian Gaming

    Science.gov (United States)

    2012-07-12

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes approval by the Department of an extension to the Class III Gaming Compact between the State of California...

  8. 77 FR 59641 - Indian Gaming

    Science.gov (United States)

    2012-09-28

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes an extension of Gaming between the Rosebud Sioux Tribe and the State of South Dakota. DATES...

  9. 78 FR 17428 - Indian Gaming

    Science.gov (United States)

    2013-03-21

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes the approval of the Class III Tribal- State Gaming Compact between the Pyramid Lake Paiute Tribe and...

  10. 78 FR 26801 - Indian Gaming

    Science.gov (United States)

    2013-05-08

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs [DR.5B711.IA000813] Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes the approval of an amendment to the Class III Tribal-State Gaming Compact...

  11. 78 FR 62650 - Indian Gaming

    Science.gov (United States)

    2013-10-22

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs [DR.5B711.IA000813] Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of extension of Tribal-State Class III Gaming Compact. SUMMARY: This publishes notice of the extension of the Class III gaming compact between the Rosebud Sioux...

  12. 78 FR 54908 - Indian Gaming

    Science.gov (United States)

    2013-09-06

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs [DR.5B711.IA000813] Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes the approval of the Class III Tribal- State Gaming Compact between the...

  13. 78 FR 62649 - Indian Gaming

    Science.gov (United States)

    2013-10-22

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs [DR.5B711.IA000813] Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Tribal-State Class III Gaming Compact taking effect. SUMMARY: This notice publishes the Class III Gaming Compact between the North Fork Rancheria of Mono...

  14. 76 FR 52968 - Indian Gaming

    Science.gov (United States)

    2011-08-24

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an extension of Gaming between the Rosebud Sioux Tribe and the State of South Dakota. DATES...

  15. 78 FR 78377 - Indian Gaming

    Science.gov (United States)

    2013-12-26

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs [DR.5B711.IA000814] Indian Gaming AGENCY... Gaming Compact. SUMMARY: This publishes notice of the extension of the Class III gaming compact between... FURTHER INFORMATION CONTACT: Paula L. Hart, Director, Office of Indian Gaming, Office of the Deputy...

  16. 76 FR 33341 - Indian Gaming

    Science.gov (United States)

    2011-06-08

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal--State Class III Gaming Compact. SUMMARY: This notice publishes an extension of Gaming between the Rosebud Sioux Tribe and the State of South Dakota. DATES...

  17. 75 FR 55823 - Indian Gaming

    Science.gov (United States)

    2010-09-14

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes an extension of Gaming between the Oglala Sioux Tribe and the State of South Dakota. DATES: Effective...

  18. 78 FR 44146 - Indian Gaming

    Science.gov (United States)

    2013-07-23

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Tribal-State Class III Gaming Compact taking effect. SUMMARY: This notice publishes the Class III Amended and Restated Tribal-State Gaming Compact between the Shingle Springs Band of...

  19. 78 FR 54670 - Indian Gaming

    Science.gov (United States)

    2013-09-05

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs [DR.5B711.IA000813] Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of extension of Tribal--State Class III Gaming Compact. SUMMARY: This publishes notice of the Extension of the Class III gaming compact between the Yankton Sioux...

  20. 78 FR 33435 - Indian Gaming

    Science.gov (United States)

    2013-06-04

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Amendments. SUMMARY: This notice publishes approval of an Agreement to Amend the Class III Tribal-State Gaming Compact between the Salt River...

  1. 78 FR 17427 - Indian Gaming

    Science.gov (United States)

    2013-03-21

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes... Gaming (Compact). DATES: Effective Date: March 21, 2013. FOR FURTHER INFORMATION CONTACT: Paula L. Hart...

  2. 78 FR 11221 - Indian Gaming

    Science.gov (United States)

    2013-02-15

    ... DEPARTMENT OF THE INTERIOR Bureau of Indian Affairs Indian Gaming AGENCY: Bureau of Indian Affairs, Interior. ACTION: Notice of Approved Tribal-State Class III Gaming Compact. SUMMARY: This notice publishes an extension of the gaming compact between the Oglala Sioux Tribe and the State of South Dakota...

  3. Facts about American Indian Education

    Science.gov (United States)

    American Indian College Fund, 2010

    2010-01-01

    As a result of living in remote rural areas, American Indians living on reservations have limited access to higher education. One-third of American Indians live on reservations, according to the U.S. Census Bureau. According to the most recent U.S. government statistics, the overall poverty rate for American Indians/Alaska Natives, including…

  4. Leadership Challenges in Indian Country.

    Science.gov (United States)

    Horse, Perry

    2002-01-01

    American Indian leaders must meld the holistic and cyclical world view of Indian peoples with the linear, rational world view of mainstream society. Tribal leaders need to be statesmen and ethical politicians. Economic and educational development must be based on disciplined long-range planning and a strong, Indian-controlled educational base.…

  5. Final Techno-Economic Analysis of 550 MWe Supercritical PC Power Plant CO2 Capture with Linde-BASF Advanced PCC Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bostick, Devin [Linde LLC, Murray Hill, NJ (United States); Stoffregen, Torsten [Linde AG Linde Engineering Division, Dresden (Germany); Rigby, Sean [BASF Corporation, Houston, TX (United States)

    2017-01-09

    This topical report presents the techno-economic evaluation of a 550 MWe supercritical pulverized coal (PC) power plant utilizing Illinois No. 6 coal as fuel, integrated with 1) a previously presented (for a subcritical PC plant) Linde-BASF post-combustion CO2 capture (PCC) plant incorporating BASF’s OASE® blue aqueous amine-based solvent (LB1) [Ref. 6] and 2) a new Linde-BASF PCC plant incorporating the same BASF OASE® blue solvent that features an advanced stripper interstage heater design (SIH) to optimize heat recovery in the PCC process. The process simulation and modeling for this report is performed using Aspen Plus V8.8. Technical information from the PCC plant is determined using BASF’s proprietary thermodynamic and process simulation models. The simulations developed and resulting cost estimates are first validated by reproducing the results of DOE/NETL Case 12 representing a 550 MWe supercritical PC-fired power plant with PCC incorporating a monoethanolamine (MEA) solvent as used in the DOE/NETL Case 12 reference [Ref. 2]. The results of the techno-economic assessment are shown comparing two specific options utilizing the BASF OASE® blue solvent technology (LB1 and SIH) to the DOE/NETL Case 12 reference. The results are shown comparing the energy demand for PCC, the incremental fuel requirement, and the net higher heating value (HHV) efficiency of the PC power plant integrated with the PCC plant. A comparison of the capital costs for each PCC plant configuration corresponding to a net 550 MWe power generation is also presented. Lastly, a cost of electricity (COE) and cost of CO2 captured assessment is shown illustrating the substantial cost reductions achieved with the Linde-BASF PCC plant utilizing the advanced SIH configuration in combination with BASF’s OASE® blue solvent technology as compared to the DOE/NETL Case 12 reference. The key factors contributing to the reduction of COE and the cost of CO2 captured

  6. The Living Indian Critical Tradition

    Directory of Open Access Journals (Sweden)

    Vivek Kumar Dwivedi

    2010-11-01

    Full Text Available This paper attempts to establish the identity of something that is often considered to be missing – a living Indian critical tradition. I refer to the tradition that arises out of the work of those Indians who write in English. The chief architects of this tradition are Sri Aurobindo, C.D. Narasimhaiah, Gayatri Chakravorty Spivak and Homi K. Bhabha. It is possible to believe that Indian literary theories derive almost solely from ancient Sanskrit poetics. Or, alternatively, one can be concerned about the sad state of affairs regarding Indian literary theories or criticism in English. There have been scholars who have raised the question of the pathetic state of Indian scholarship in English and have even come up with some positive suggestions. But these scholars are those who are ignorant about the living Indian critical tradition. The significance of the Indian critical tradition lies in the fact that it provides the real focus to the Indian critical scene. Without an awareness of this tradition Indian literary scholarship (which is quite a different thing from Indian literary criticism and theory as it does not have the same impact as the latter two do can easily fail to see who the real Indian literary critics and theorists are.

  7. Indian Women: An Historical and Personal Perspective

    Science.gov (United States)

    Christensen, Rosemary Ackley

    1975-01-01

    Several issues relating to Indian women are discussed. These include (1) the three types of people to whom we owe our historical perceptions of Indian women, (2) role delineation in Indian society; (3) differences between Indian women and white women, and (4) literary role models of Indian women. (Author/BW)

  8. A promising niche: waste to energy project in the Indian dairy sector

    International Nuclear Information System (INIS)

    Patankar, Mahesh; Patwardhan, Anand; Verbong, Geert

    2010-01-01

    The dairy sector is known to have significant local and global environmental impacts; but it also has proven renewable-energy generation potential. This paper analyzes a specific niche experiment in the Indian dairy industry, wherein cattle waste management is carried out by a multitude of stakeholders. These include the waste collectors, local technology adopters, research institutions, multilateral donor agencies, the Indian government, technology suppliers and operation and maintenance teams who have managed an uninterrupted 1 MWe energy production over the past 4 years. This analysis uses the sociotechnical regime framework to study the interaction of social, technological, economic and policy-related aspects relevant to the niche experiment. The analysis shows a potential to contribute to the development of two complementing regimes-one related to cooperative waste management and the other related to grid-connected renewable-energy-based electricity generation. Key factors for a successful development are not only a long-term financing protection through government subsidies to cover higher capital cost and a preferential tariff related to energy throughput, but also the adaptation of technology, the embedding in the local cooperative structure and the removal of regulatory barriers.

  9. INDIAN ACADEMY OF SCIENCES

    Indian Academy of Sciences (India)

    user

    2016-07-02

    Jul 2, 2016 ... P R O G R A M M E. 1 July 2016 (Friday). Venue: Faculty Hall, Indian Institute of Science, Bengaluru ... 1800–1900 Session 1E – Public Lecture. Pratap Bhanu Mehta, Centre for Policy Research, New Delhi. Two ideas of India.

  10. Indian Astronomy: History of

    Science.gov (United States)

    Mercier, R.; Murdin, P.

    2002-01-01

    From the time of A macronryabhat under dota (ca AD 500) there appeared in India a series of Sanskrit treatises on astronomy. Written always in verse, and normally accompanied by prose commentaries, these served to create an Indian tradition of mathematical astronomy which continued into the 18th century. There are as well texts from earlier centuries, grouped under the name Jyotishaveda macronn d...

  11. The Indian Monsoon

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 13; Issue 3. The Indian Monsoon - Links to Cloud systems over the Tropical Oceans. Sulochana Gadgil. Series Article Volume 13 Issue 3 March 2008 pp 218-235. Fulltext. Click here to view fulltext PDF. Permanent link:

  12. Becoming an Indian

    Indian Academy of Sciences (India)

    Ramachandra Guha

    2017-11-25

    Nov 25, 2017 ... learning science by what he later recalled as 'Gandhian or basic .... Calcutta to offer their thoughts on Indian planning. Hal- ... had come to India for good. But any .... am eager to be of help and service to a sincere soul like you.

  13. Indians of North Carolina.

    Science.gov (United States)

    Bureau of Indian Affairs (Dept. of Interior), Washington, DC.

    Published by the U.S. Department of the Interior, this brief booklet on the historical development of the Cherokee Nation emphasizes the Tribe's relationship with the Bureau of Indian Affairs and its improved economy. Citing tourism as the major tribal industry, tribal enterprises are named and described (a 61 unit motor court in existence since…

  14. Indian Health Disparities

    Science.gov (United States)

    ... reservations and in rural communities, mostly in the western United States and Alaska. The American Indian and ... Office of Finance and Accounting - 10E54 Office of Human Resources - 11E53A Office of Information Technology - 07E57B Office of ...

  15. Caregiving in Indian Country

    Centers for Disease Control (CDC) Podcasts

    2009-12-23

    This podcast discusses the role of caregivers in Indian County and the importance of protecting their health. It is primarily targeted to public health and aging services professionals.  Created: 12/23/2009 by National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP).   Date Released: 12/23/2009.

  16. Depreciation of the Indian Currency: Implications for the Indian Economy.

    OpenAIRE

    Sumanjeet Singh

    2009-01-01

    The Indian currency has depreciated by more than 20 per cent since April 2008 and breached its crucial 50-level against the greenback on sustained dollar purchases by foreign banks and stronger dollar overseas. The fall in the value of Indian rupee has several consequences which could have mixed effects on Indian economy. But, mainly, there are four expected implications of falling rupee. First, it should boost exports; second, it will lead to higher cost of imported goods and make some of th...

  17. New fellows | Announcements | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    ... of Medical Sciences, New Delhi; S K Bhowmik, Indian Institute of Technology, ... Souvik Mahapatra, Indian Institute of Technology, Mumbai; Prabal K Maiti, Indian ... Math Art and Design: MAD about Math, Math Education and Outreach.

  18. Asthma and American Indians/Alaska Natives

    Science.gov (United States)

    ... Minority Population Profiles > American Indian/Alaska Native > Asthma Asthma and American Indians/Alaska Natives In 2015, 240, ... Native American adults reported that they currently have asthma. American Indian/Alaska Native children are 60% more ...

  19. Influence of the overfire air ratio on the NO(x) emission and combustion characteristics of a down-fired 300-MW(e) utility boiler.

    Science.gov (United States)

    Ren, Feng; Li, Zhengqi; Chen, Zhichao; Fan, Subo; Liu, Guangkui

    2010-08-15

    Down-fired boilers used to burn low-volatile coals have high NO(x) emissions. To find a way of solving this problem, an overfire air (OFA) system was introduced on a 300 MW(e) down-fired boiler. Full-scale experiments were performed on this retrofitted boiler to explore the influence of the OFA ratio (the mass flux ratio of OFA to the total combustion air) on the combustion and NO(x) emission characteristics in the furnace. Measurements were taken of gas temperature distributions along the primary air and coal mixture flows, average gas temperatures along the furnace height, concentrations of gases such as O(2), CO, and NO(x) in the near-wall region and carbon content in the fly ash. Data were compared for five different OFA ratios. The results show that as the OFA ratio increases from 12% to 35%, the NO(x) emission decreases from 1308 to 966 mg/Nm(3) (at 6% O(2) dry) and the carbon content in the fly ash increases from 6.53% to 15.86%. Considering both the environmental and economic effect, 25% was chosen as the optimized OFA ratio.

  20. Improving combustion characteristics and NO(x) emissions of a down-fired 350 MW(e) utility boiler with multiple injection and multiple staging.

    Science.gov (United States)

    Kuang, Min; Li, Zhengqi; Xu, Shantian; Zhu, Qunyi

    2011-04-15

    Within a Mitsui Babcock Energy Limited down-fired pulverized-coal 350 MW(e) utility boiler, in situ experiments were performed, with measurements taken of gas temperatures in the burner and near the right-wall regions, and of gas concentrations (O(2) and NO) from the near-wall region. Large combustion differences between zones near the front and rear walls and particularly high NO(x) emissions were found in the boiler. With focus on minimizing these problems, a new technology based on multiple-injection and multiple-staging has been developed. Combustion improvements and NO(x) reductions were validated by investigating three aspects. First, numerical simulations of the pulverized-coal combustion process and NO(x) emissions were compared in both the original and new technologies. Good agreement was found between simulations and in situ measurements with the original technology. Second, with the new technology, gas temperature and concentration distributions were found to be symmetric near the front and rear walls. A relatively low-temperature and high-oxygen-concentration zone formed in the near-wall region that helps mitigate slagging in the lower furnace. Third, NO(x) emissions were found to have decreased by as much as 50%, yielding a slight decrease in the levels of unburnt carbon in the fly ash.

  1. Reducing NOx Emissions for a 600 MWe Down-Fired Pulverized-Coal Utility Boiler by Applying a Novel Combustion System.

    Science.gov (United States)

    Ma, Lun; Fang, Qingyan; Lv, Dangzhen; Zhang, Cheng; Chen, Yiping; Chen, Gang; Duan, Xuenong; Wang, Xihuan

    2015-11-03

    A novel combustion system was applied to a 600 MWe Foster Wheeler (FW) down-fired pulverized-coal utility boiler to solve high NOx emissions, without causing an obvious increase in the carbon content of fly ash. The unit included moving fuel-lean nozzles from the arches to the front/rear walls and rearranging staged air as well as introducing separated overfire air (SOFA). Numerical simulations were carried out under the original and novel combustion systems to evaluate the performance of combustion and NOx emissions in the furnace. The simulated results were found to be in good agreement with the in situ measurements. The novel combustion system enlarged the recirculation zones below the arches, thereby strengthening the combustion stability considerably. The coal/air downward penetration depth was markedly extended, and the pulverized-coal travel path in the lower furnace significantly increased, which contributed to the burnout degree. The introduction of SOFA resulted in a low-oxygen and strong-reducing atmosphere in the lower furnace region to reduce NOx emissions evidently. The industrial measurements showed that NOx emissions at full load decreased significantly by 50%, from 1501 mg/m3 (O2 at 6%) to 751 mg/m3 (O2 at 6%). The carbon content in the fly ash increased only slightly, from 4.13 to 4.30%.

  2. Thermo-mechanical design and structural analysis of the first wall for ARIES-III, A 1000 MWeD-3He power reactor

    International Nuclear Information System (INIS)

    Sviatoslavsky, I.; Blanchard, J.P.; Mogahed, E.A.

    1992-01-01

    This paper reports on ARIES III, a conceptual design study of a 1000 MWe D- 3 He tokamak fusion power reactor in which most of the energy comes from charged particle transport, bremsstrahlung and synchrotron radiation, and only a small fraction (∼ 4%) comes form neutrons. This form of energy is deposited as surface heating on the chamber first wall (FW) and divertor elements, while the neutron energy is deposited as bulk nuclear heating within the shield. Since this reactor does not use tritium, there is no breeding blanket. Instead a shield is provided to protect the magnets from neutrons. The Fw is very unique in a D- 3 He reactor, it must be capable of absorbing the high surface heat in a mode suitable for efficient power cycle conversion, it must be able to reflect synchrotron radiation, and it must be able to withstand high current plasma disruptions. The FW is made of a low activation ferritic steel (MHT-9) and is cooled with an organic coolant (HB-40) at a pressure of 2 MPa. The FW has a coating of 0.01 cm tungsten on the MHT-9, followed by 0.15 cm of Be on the plasma side. This is needed for synchrotron radiation reflection and as a melt layer to guard against the thermal effects of a plasma disruption

  3. Comparison of control rod effectiveness for thorium and low-enriched fuel cycles in the GA-1, 160 MW(e) design

    Energy Technology Data Exchange (ETDEWEB)

    Neef, Hans Joachim

    1974-03-15

    In an investigation of the properties of the Thorium-Uranium (Th) and the Low-Enriched Uranium (LEU) fuel cycles it is also necessary to compare the effectiveness of the control rods in a reactor system operating with these sorts of fuel. Furthermore, it is under consideration to start a reactor with LEU fuel and switch-over to a Th cycle. It is also of interest to look at the switch-over phase in respect to the control rod effectiveness. The various fuel cycles have been studied for the same fuel element and control rod design, namely the one of GA's commercially available 1,160 MW(e) reference power station. This paper gives the first results on the control rod calculations and is presented mainly in two parts. Part 1 describes spectral effects which have been investigated by cell calculations with a discrete ordinates transport code. The main result is the higher effectiveness of a rod in a Th-cycle compared with a LEU-cycle. Part 2 reports on reactor calculations with a diffusion code and shows that this advantage can partially disappear in the reactor because of the spatial flux distribution. This effect has to be studied in further investigations for a full understanding.

  4. Conceptual design considerations for providing hook-up type schemes for tracking beyond design basis events (BDBE) for 700 MWe PHWR project

    International Nuclear Information System (INIS)

    Vhora, S.F.; Inder Jit; Bhardwaj, S.A.

    2005-01-01

    A broad review of major nuclear accidents such as Chernobyl reveals that provision of access to the reactor core for cooling purpose had to be made from outside the reactor building by tunneling. Also the NAPS fire incident could be mitigated once the fire water injection to the steam generators could be ensured. In this case the boiler room which was outside the primary containment was accessible relatively easily for mitigation after the initial period. Both of the above had accident scenarios which can be termed Beyond Design Basis (BDBE) since the accident initiation/scenario did not fit into the events under postulated initiating events (PIES) or Design Basis Events (DBEs). These accidents or events reveal that some sort of access to the core or the components inside the Reactor building becomes necessary. It is also to be noted that manual intervention beyond the initial period of half an hour or earlier in the Emergency operating procedure (EOP) is inevitably called for as a recovery action in order to mitigate the severity and minimize long term consequences. This paper attempts to discuss the type of concepts which can give access to the core or associated systems which can then provide continued heat sink. The discussions would include the criteria for design of such concepts and give examples of such concepts already implemented and proposes schemes to be implemented in the 700 MWe Project. (author)

  5. The impact of the CO_2 separation system integration with a 900 MW_e power unit on its thermodynamic and economic indices

    International Nuclear Information System (INIS)

    Łukowicz, Henryk; Mroncz, Marcin

    2015-01-01

    This paper presents an analysis of the thermal cycle of a supercritical 900 MW_e condensing power plant which meets the “capture ready” requirements. The CO_2 separation method selected for the analysis is chemical absorption using MEA (monoethanolamine) or ammonia as sorbent. The indispensable scope of the turbine system upgrade necessitated by the incorporation of the carbon dioxide separation installation is proposed. The change in indices of the power unit operation after integration with the capture installation is presented for different variants of the retrofit. If MEA is used for carbon dioxide separation, the smallest drop in electric power can be observed in the case of hard coal for added stages at the intermediate pressure part outlet. In the case of lignite, the most favourable upgrade solution in terms of the smallest drop in electric power is elimination of one low pressure part and a backpressure turbine is added at the same time. If ammonia is used as sorbent, the best upgrade solution in terms of the smallest drop in electric power is the variant with more stages added at the IP part outlet, regardless of the type of fuel. An economic analysis is conducted for the proposed variants. - Highlights: • Impact of steam extraction on the turbine operation. • Adding more stages at the intermediate pressure part outlet. • Installing a backpressure turbine. • Eliminating one of the operating low pressure parts. • Economic assessment of proposed variants.

  6. Study of the muon spectrum at a depth 570 m.w.e. underground by means of the 100-ton scintillation detector

    International Nuclear Information System (INIS)

    Enikeev, R.I.; Zatsepin, G.T.; Korol'kova, E.V.; Kudryavtsev, V.A.; Mal'gin, A.S.; Ryazhskaya, O.G.; Khal'chukov, F.F.

    1988-01-01

    The experiment was carried out using the 100-ton apparatus at the Artemovsk Scientific Station of the Institute of Nuclear Research, USSR Academy of Sciences, located in a salt mine at a depth 570 m.w.e. underground. The spectrum of the energy release in the electromagnetic cascades which are generated by muons underground was measured. The electromagnetic and nuclear cascades were separated on the basis of the number of neutrons in these cascades. The spectrum of the energy release obtained is consistent with a spectrum of π and K mesons with γ/sub π//sub ,//sub K/ = 1.75 +- 0.08 for muon energies at sea level E 0 /sub μ/ >0.7 TeV. The experimental data recalculated to the vertical spectrum of muons at sea level agree with the results of other studies. Up to energies of about 100 TeV neither the spectrum of the primary cosmic rays nor the characteristics of the pA interaction undergo changes which could lead to an increase of γ/sub π//sub ,//sub K/ to a value exceeding 1.85

  7. Safety performance comparation of MOX, nitride and metallic fuel based 25-100 MWe Pb-Bi cooled long life fast reactors without on-site refuelling

    International Nuclear Information System (INIS)

    Su'ud, Zaki

    2008-01-01

    In this paper the safety performance of 25-100 MWe Pb-Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb-Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance. The results of safety analysis of long life Pb-Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores. (author)

  8. Hollow Fiber Membrane Contactors for CO2 Capture: Modeling and Up-Scaling to CO2 Capture for an 800 MWe Coal Power Station

    Directory of Open Access Journals (Sweden)

    Kimball Erin

    2014-11-01

    Full Text Available A techno-economic analysis was completed to compare the use of Hollow Fiber Membrane Modules (HFMM with the more conventional structured packing columns as the absorber in amine-based CO2 capture systems for power plants. In order to simulate the operation of industrial scale HFMM systems, a two-dimensional model was developed and validated based on results of a laboratory scale HFMM. After successful experiments and validation of the model, a pilot scale HFMM was constructed and simulated with the same model. The results of the simulations, from both sizes of HFMM, were used to assess the feasibility of further up-scaling to a HFMM system to capture the CO2 from an 800 MWe power plant. The system requirements – membrane fiber length, total contact surface area, and module volume – were determined from simulations and used for an economic comparison with structured packing columns. Results showed that a significant cost reduction of at least 50% is required to make HFMM competitive with structured packing columns. Several factors for the design of industrial scale HFMM require further investigation, such as the optimal aspect ratio (module length/diameter, membrane lifetime, and casing material and shape, in addition to the need to reduce the overall cost. However, HFMM were also shown to have the advantages of having a higher contact surface area per unit volume and modular scale-up, key factors for applications requiring limited footprints or flexibility in configuration.

  9. Solar thermal power stations for activities implemented jointly. The Theseus 50 MWe solar thermal power plant for the island of Crete, Greece

    Energy Technology Data Exchange (ETDEWEB)

    Brakmann, Georg [Fichtner, Stuttgart (Germany); Aringhoff, Rainer [Pilkington Solar International (United Kingdom); Cobi, Arend [PreussenElektra (Germany)

    1998-09-01

    THESEUS, the proposed commercial 50 MWe (net) Thermal Solar European Power Station for the Island of Crete is a solar hybrid plant with parabolic trough collectors and an advanced high efficiency Rankine reheat steam cycle. At the end of 1996 the DG XVII (Energy) of the European Commission has accepted the THERMIE application of the THESEUS consortium for the design phase. THESEUS reduces the required oil imports by 28 000 t/a, thereby saving the Greek economy every year 4 million ECU in foreign currency. During its 25 years technical lifetime 2.2 million tons of CO{sub 2} emissions will be avoided. Supply, construction, erection and operation of THESEUS creates 2 000 qualified employments (man-years). Because of the high manpower intensity of solar plants and their larger capital income from interest payments in contrast to the high fuel import intensity of fossil plants, THESEUS will generate larger tax revenues for Greece and for the supplier`s countries. The investment cost of THESEUS is some 135 million ECU. Even without any subsidies this would result in electricity generation cost of some 0.085 ECY/kWh, which is lower than the current average cost from the existing power plants of Crete. (author)

  10. BIA Indian Lands Dataset (Indian Lands of the United States)

    Data.gov (United States)

    Federal Geographic Data Committee — The American Indian Reservations / Federally Recognized Tribal Entities dataset depicts feature location, selected demographics and other associated data for the 561...

  11. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  12. Celebrating National American Indian Heritage Month

    National Research Council Canada - National Science Library

    Mann, Diane

    2004-01-01

    November has been designated National American Indian Heritage Month to honor American Indians and Alaska Natives by increasing awareness of their culture, history, and, especially, their tremendous...

  13. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Last known address: Professor, Department of Chemistry, Indian Institute of ... Specialization: Natural Products & Drug Development, Reaction Mechanism, ... Specialization: Plant Molecular Biology, Plant Tissue Culture and Genetic ...

  14. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Department of Electrical Engineering, Indian Institute of Technology, Powai, Mumbai ..... Specialization: Elementary Particle Physics ..... Sciences, National Institute of Science Education & Research, Jatni, Khordha 752 050, Orissa

  15. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Specialization: DNA Double-Strand Break Repair, Genomic Instability, Cancer ... Address: Indian Institute of Science Education & Research, Dr Homi Bhabha Road, .... Inflammatory Bowel Disease, Gastrointestinal Microbiome Stem Cells

  16. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Time Programs, Logic Programs, Mobile Computing and Computer & Information Security Address: Distinguished V Professor, Computer Science & Engineering Department, Indian Institute of Technology, Powai, Mumbai 400 076, Maharashtra

  17. Indian Danish intermarriage

    DEFF Research Database (Denmark)

    Singla, Rashmi; Sriram, Sujata

    This paper explores motivations of Indian partner in mixed Indian-Danish couples living in Denmark. One of the characteristics of modernity is increased movements across borders, leading to increased intimate relationships across national/ethnic borders. The main research question here deals...... with the reasons for couple ‘getting together’. How do motives interplay with the gender- and the family generational, socio -economical categories? The paper draws from an explorative study conducted in Denmark among intermarried couples, consisting of in-depth interviews with ten ‘ordinary’ intermarried couples...... (TEM), transnationalism and a phenomenological approach to sexual desire and love. We find that there are three different pathways, highlighting commonality of work identity, a cosmopolitan identity and academic interests, where differential changing patterns of privileges and power are also evoked...

  18. Indian President visits CERN

    CERN Multimedia

    Katarina Anthony

    2011-01-01

    On 1 October, her Excellency Mrs Pratibha Devisingh Patil, President of India, picked CERN as the first stop on her official state visit to Switzerland. Accompanied by a host of Indian journalists, a security team, and a group of presidential delegates, the president left quite an impression when she visited CERN’s Point 2!   Upon arrival, Pratibha Patil was greeted by CERN Director General Rolf Heuer, as well as senior Indian scientists working at CERN, and various department directors. After a quick overview of the Organization, Rolf Heuer and the President addressed India’s future collaboration with CERN. India is currently an Observer State of the Organization, and is considering becoming an Associate Member State. A short stop in LHC operations gave Steve Myers and the Accelerator team the opportunity to take the President on a tour through the LHC tunnel. From there, ALICE’s Tapan Nayak and Spokesperson Paolo Giubellino took Pratibha Patil to the experiment&am...

  19. Indian cosmogonies and cosmologies

    Directory of Open Access Journals (Sweden)

    Pajin Dušan

    2011-01-01

    Full Text Available Various ideas on how the universe appeared and develops, were in Indian tradition related to mythic, religious, or philosophical ideas and contexts, and developed during some 3.000 years - from the time of Vedas, to Puranas. Conserning its appeareance, two main ideas were presented. In one concept it appeared out of itself (auto-generated, and gods were among the first to appear in the cosmic sequences. In the other, it was a kind of divine creation, with hard work (like the dismembering of the primal Purusha, or as emanation of divine dance. Indian tradition had also various critiques of mythic and religious concepts (from the 8th c. BC, to the 6c., who favoured naturalistic and materialistic explanations, and concepts, in their cosmogony and cosmology. One the peculiarities was that indian cosmogony and cosmology includes great time spans, since they used a digit system which was later (in the 13th c. introduced to Europe by Fibonacci (Leonardo of Pisa, 1170-1240.

  20. Working Women: Indian Perspective

    Directory of Open Access Journals (Sweden)

    Dharmendra MEHTA

    2014-06-01

    Full Text Available In India, due to unprecedented rise in the cost of living, ris-ing prices of commodities, growing expenses on children ed-ucation, huge rate of unemployment, and increasing cost of housing properties compel every Indian family to explore all the possible ways and means to increase the household income. It is also witnessed that after globalization Indian women are able to get more jobs but the work they get is more casual in nature or is the one that men do not prefer to do or is left by them to move to higher or better jobs. Working women refers to those in paid employment. They work as lawyers, nurses, doctors, teachers and secretaries etc. There is no profession today where women are not employed. University of Oxford’s Professor Linda Scott recently coined the term the Double X Economy to describe the global economy of women. The present paper makes an attempt to discuss issues and challenges that are being faced by Indian working women at their respective workstations.

  1. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. PRIYANKA SHUKLA. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 133-143 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Grad-type fourteen-moment theory for ...

  2. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. SERGEY P KUZNETSOV. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 117-132 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Chaos in three coupled rotators: ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. NORBERT MARWAN. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 51-60 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Inferring interdependencies from short time ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. GIOVANNA ZIMATORE. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 35-41 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. RQA correlations on real business cycles ...

  5. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. SUDHARSANA V IYENGAR. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 93-99 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Missing cycles: Effect of climate ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. BEDARTHA GOSWAMI. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 51-60 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Inferring interdependencies from short ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. MURILO S BAPTISTA. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 17-23 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Interpreting physical flows in networks as a ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. F REVUELTA. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 145-155 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Rate calculation in two-dimensional barriers with ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. JOYDEEP SINGHA. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 195-203 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Spatial splay states in coupled map lattices ...

  10. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. F FAMILY. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 221-224 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Transport in ratchets with single-file constraint.

  11. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. JANAKI BALAKRISHNAN. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 93-99 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Missing cycles: Effect of climate change ...

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    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series. PAUL SCHULTZ. Articles written in Indian Academy of Sciences Conference Series. Volume 1 Issue 1 December 2017 pp 51-60 Proceedings of the Conference on Perspectives in Nonlinear Dynamics - 2016. Inferring interdependencies from short time ...

  13. Overall evaluation of combustion and NO(x) emissions for a down-fired 600 MW(e) supercritical boiler with multiple injection and multiple staging.

    Science.gov (United States)

    Kuang, Min; Li, Zhengqi; Liu, Chunlong; Zhu, Qunyi

    2013-05-07

    To achieve significant reductions in NOx emissions and to eliminate strongly asymmetric combustion found in down-fired boilers, a deep-air-staging combustion technology was trialed in a down-fired 600 MWe supercritical utility boiler. By performing industrial-sized measurements taken of gas temperatures and species concentrations in the near wing-wall region, carbon in fly ash and NOx emissions at various settings, effects of overfire air (OFA) and staged-air damper openings on combustion characteristics, and NOx emissions within the furnace were experimentally determined. With increasing the OFA damper opening, both fluctuations in NOx emissions and carbon in fly ash were initially slightly over OFA damper openings of 0-40% but then lengthened dramatically in openings of 40-70% (i.e., NOx emissions reduced sharply accompanied by an apparent increase in carbon in fly ash). Decreasing the staged-air declination angle clearly increased the combustible loss but slightly influenced NOx emissions. In comparison with OFA, the staged-air influence on combustion and NOx emissions was clearly weaker. Only at a high OFA damper opening of 50%, the staged-air effect was relatively clear, i.e., enlarging the staged-air damper opening decreased carbon in fly ash and slightly raised NOx emissions. By sharply opening the OFA damper to deepen the air-staging conditions, although NOx emissions could finally reduce to 503 mg/m(3) at 6% O2 (i.e., an ultralow NOx level for down-fired furnaces), carbon in fly ash jumped sharply to 15.10%. For economical and environment-friendly boiler operations, an optimal damper opening combination (i.e., 60%, 50%, and 50% for secondary air, staged-air, and OFA damper openings, respectively) was recommended for the furnace, at which carbon in fly ash and NOx emissions attained levels of about 10% and 850 mg/m(3) at 6% O2, respectively.

  14. Joint studies of LOF and TOP incidents for a 1300 MW(E) LMFBR using the computer codes SAS3D/EPIC and FRAX-2

    International Nuclear Information System (INIS)

    Leslie, R.; Billington, D.E.; Mann, J.E.

    1982-04-01

    The results of joint studies carried out for a 1300MW(E) LMFBR are described. The incidents examined were a slow TOP (3c/s) and a LOF (pump rundown with 9s flow halving time), both with failure to trip. For the TOP incident a benign outcome was predicted largely as a consequence of the prediction of clad failure near the top of the core. For the LOF incident highly energetic outcomes were not predicted for the reference case because the incident was terminated by disassembly (by fuel vapour pressure) in voided channels and failures in low-rated flooded channels with MFCI potential were not predicted. In the variant cases where MFCIs were predicted before shutdown, and rapid enough extension of the clad rips was allowed, low energetics were still predicted as a consequence of fuel sweepout. The strength of the MFCIs (as represented by a Cho-Wright treatment) does not appear to be an important factor but the results are dependent on the prediction of negative reactivity addition through fuel sweepout. The physical conditions obtaining at the time of fuel failure are such as to suggest that internal fuel motion following failure should not have an important effect on accident energetics, unless the development of the initial rip is delayed by several milliseconds. This is an area where only limited experimental evidence is available. Other areas of uncertainty are associated with the position of failure, of clad rip propagation and the influence of incoherency on the progression of the incident. Clad motion effects were shown not to influence accident energetics significantly for the reactor model considered. (author)

  15. Assessment on 900–1300 MWe PWRs of the ASTEC-based simulation tool of SGTR thermal-hydraulics for the IRSN Emergency Technical Centre

    Energy Technology Data Exchange (ETDEWEB)

    Foucher, L., E-mail: laurent.foucher@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Cousin, F.; Fleurot, J. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SAG, Cadarache, Saint-Paul-lez-Durance 13115 (France); Brethes, S. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PRP-CRI/SESUC, Cadarache, Saint-Paul-lez-Durance 13115 (France)

    2014-06-01

    In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc. In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool. This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool.

  16. Measurement of gas species, temperatures, coal burnout, and wall heat fluxes in a 200 MWe lignite-fired boiler with different overfire air damper openings

    Energy Technology Data Exchange (ETDEWEB)

    Jianping Jing; Zhengqi Li; Guangkui Liu; Zhichao Chen; Chunlong Liu [Harbin Institute of Technology, Harbin (China). School of Energy Science and Engineering

    2009-07-15

    Measurements were performed on a 200 MWe, wall-fired, lignite utility boiler. For different overfire air (OFA) damper openings, the gas temperature, gas species concentration, coal burnout, release rates of components (C, H, and N), furnace temperature, and heat flux and boiler efficiency were measured. Cold air experiments for a single burner were conducted in the laboratory. The double-swirl flow pulverized-coal burner has two ring recirculation zones starting in the secondary air region in the burner. As the secondary air flow increases, the axial velocity of air flow increases, the maxima of radial velocity, tangential velocity and turbulence intensity all increase, and the swirl intensity of air flow and the size of recirculation zones increase slightly. In the central region of the burner, as the OFA damper opening widens, the gas temperature and CO concentration increase, while the O{sub 2} concentration, NOx concentration, coal burnout, and release rates of components (C, H, and N) decrease, and coal particles ignite earlier. In the secondary air region of the burner, the O{sub 2} concentration, NOx concentration, coal burnout, and release rates of components (C, H, and N) decrease, and the gas temperature and CO concentration vary slightly. In the sidewall region, the gas temperature, O{sub 2} concentration, and NOx concentration decrease, while the CO concentration increases and the gas temperature varies slightly. The furnace temperature and heat flux in the main burning region decrease appreciably, but increase slightly in the burnout region. The NOx emission decreases from 1203.6 mg/m{sup 3} (6% O{sub 2}) for a damper opening of 0% to 511.7 mg/m{sup 3} (6% O{sub 2}) for a damper opening of 80% and the boiler efficiency decreases from 92.59 to 91.9%. 15 refs., 17 figs., 3 tabs.

  17. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  18. Assessment on 900–1300 MWe PWRs of the ASTEC-based simulation tool of SGTR thermal-hydraulics for the IRSN Emergency Technical Centre

    International Nuclear Information System (INIS)

    Foucher, L.; Cousin, F.; Fleurot, J.; Brethes, S.

    2014-01-01

    In the event of an accident occurring in a nuclear power plant (NPP), being able to predict the amount of released radioactive substances in the environment is of prime importance. Depending on the severity of the accident, it can be necessary to quickly and efficiently protect the population and the surrounding environment from the associated radiological consequences. In France, the IRSN Emergency Technical Centre provides a technical support in decision making in case of a nuclear accident. The main objectives are to evaluate and predict the plant behaviour and radioactive releases during the accident. Different types of complementary tools are used: expert assessments, pre-calculated databases, simulation tools, etc. In the case of Steam Generator Tube Rupture (SGTR) accidents that may lead to significant radioactive releases to the atmosphere through the steam generator relief valves, IRSN is currently improving the simulation tools for diagnosis in crisis management. The objective is to adapt the thermal-hydraulic and FP behaviour modules of the severe accident integral code ASTEC V2.0, jointly developed by IRSN and its German counterpart GRS, to crisis management requirements. These requirements impose a fast running, highly reliable (accurate physical results), flexible and simple tool. This paper summarizes the results of the benchmarks between the ASTEC V2.0 thermal-hydraulic module and the CATHARE 2 (V2.5) French reference thermal-hydraulics code on several SGTR scenarios both for PWR 900 and 1300 MWe, with a particular emphasis on the computational time and physical models assessment. The overall agreement between both codes is good on the primary and secondary circuit thermal-hydraulic parameters. Moreover, the reliability and fast computational time of the thermal-hydraulic module of ASTEC V2.0 code appeared very satisfactory and in accordance with the requirements of an emergency tool

  19. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    2018-06-07

    Jun 7, 2018 ... Science Education Programmes · Women in Science · Committee on ... Transliteration; informal information; natural language processing (NLP); information retrieval. ... Department of Computer Science and Engineering, Indian Institute of Technology (Indian School of Mines), Dhanbad 826004, India ...

  20. American Indians in Graduate Education.

    Science.gov (United States)

    Kidwell, Clara Sue

    1989-01-01

    The number of American Indians enrolled in institutions of higher education is very small. Enrollment figures for fall 1984 show Indians made up .68% of the total enrollment in institutions of higher education in the country, but only 15% of them were in universities. Their largest representation was in two-year institutions, where 54% of Indian…

  1. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Home; Journals; Sadhana. K Samudravijaya. Articles written in Sadhana. Volume 27 Issue 1 February 2002 pp 113-126. Indian accent text-to-speech system for web browsing · Aniruddha Sen K Samudravijaya · More Details Abstract Fulltext PDF. Incorporation of speech and Indian scripts can greatly enhance the ...

  2. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Department of Industrial Engineering and Management, Maulana Abul Kalam Azad University of Technology, Kolkata 700064, India; Indian Institute of Management Raipur, GEC Campus, Sejbahar, Raipur 492015, India; Indian National Centre for Ocean Information Services, Ministry of Earth Sciences, Hyderabad 500090, ...

  3. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Home; Journals; Sadhana; Volume 41; Issue 2. Nearest neighbour classification of Indian sign language gestures using kinect camera. Zafar Ahmed Ansari Gaurav Harit. Volume 41 Issue 2 February 2016 pp 161-182 ... Keywords. Indian sign language recognition; multi-class classification; gesture recognition.

  4. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Logo of the Indian Academy of Sciences. Indian Academy of ... 2013 pp 571-589. An evolutionary approach for colour constancy based on gamut mapping constraint satisfaction ... A new colour constancy algorithm based on automatic determination of gray framework parameters using neural network · Mohammad Mehdi ...

  5. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Toggle navigation. Logo of the Indian Academy of Sciences. Indian Academy of Sciences. Home · About IASc · History · Memorandum of Association ... Volume 31 Issue 5 October 2006 pp 621-633. Minimizing total costs of forest roads with computer-aided design model · Abdullah E Akay · More Details Abstract Fulltext PDF.

  6. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    2018-03-14

    Mar 14, 2018 ... Cloud security; network security; anomaly detection; network traffic analysis; DDoS attack detection. ... Department of Computer Science and Engineering, Indian Institute of Technology Roorkee, Roorkee 247667, India; Department of Applied Science and Engineering, Indian Institute of Technology ...

  7. Textbooks and the American Indian.

    Science.gov (United States)

    Costo, Rupert, Ed.

    An independent Indian publishing house has been formed to provide classroom instructional materials which deal accurately with the history, culture, and role of the American Indian. This book is a preliminary statement in that publishing program. General criteria, valid for instructional materials from elementary through high school, are applied…

  8. The average Indian female nose.

    Science.gov (United States)

    Patil, Surendra B; Kale, Satish M; Jaiswal, Sumeet; Khare, Nishant; Math, Mahantesh

    2011-12-01

    This study aimed to delineate the anthropometric measurements of the noses of young women of an Indian population and to compare them with the published ideals and average measurements for white women. This anthropometric survey included a volunteer sample of 100 young Indian women ages 18 to 35 years with Indian parents and no history of previous surgery or trauma to the nose. Standardized frontal, lateral, oblique, and basal photographs of the subjects' noses were taken, and 12 standard anthropometric measurements of the nose were determined. The results were compared with published standards for North American white women. In addition, nine nasal indices were calculated and compared with the standards for North American white women. The nose of Indian women differs significantly from the white nose. All the nasal measurements for the Indian women were found to be significantly different from those for North American white women. Seven of the nine nasal indices also differed significantly. Anthropometric analysis suggests differences between the Indian female nose and the North American white nose. Thus, a single aesthetic ideal is inadequate. Noses of Indian women are smaller and wider, with a less projected and rounded tip than the noses of white women. This study established the nasal anthropometric norms for nasal parameters, which will serve as a guide for cosmetic and reconstructive surgery in Indian women.

  9. epubworkshop | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Toggle navigation. Logo of the Indian Academy of Sciences. Indian Academy of Sciences. Home · About IASc · History · Memorandum of Association · Role of the Academy · Statutes · Council · Raman Chair · Jubilee Chair · Academy – Springer Nature chair · Academy Trust · Contact details · Office Staff · Office complaint ...

  10. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    ... features of Indian Heavy Water Reactors for prevention and mitigation of such extreme events. The probabilistic safety analysis revealed that the risk from Indian Heavy Water Reactors are negligibly small. Volume 38 Issue 6 December 2013 pp 1173-1217. Entrainment phenomenon in gas–liquid two-phase flow: A review.

  11. Home | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    2017-07-02

    Jul 2, 2017 ... The editors Biman Bagchi (FASc, FNA, FTWAS; Indian Institute of Science, Bangalore, India), David Clary (FRS; Oxford University, Oxford, UK) and N Sathyamurthy (FASc, FNA, FTWAS; Indian Institute of Science Education and Research, Mohali, India) have put together a 29 articles on theoretical physical ...

  12. Methodology for understanding Indian culture

    DEFF Research Database (Denmark)

    Sinha, Jai; Kumar, Rajesh

    2004-01-01

    Methods of understanding cultures, including Indian culture, are embedded in a broad spectrum of sociocultural approaches to human behavior in general. The approaches examined in this paper reflect evolving perspectives on Indian culture, ranging from the starkly ethnocentric to the largely...... eclectic and integrative. Most of the methods herin discussed were developed in the West and were subsequently taken up with or without adaptations to fit the Indian context. The paper begins by briefly reviewing the intrinsic concept of culture. It then adopts a historical view of the different ways...... and means by which scholars have construed the particular facets of Indian culture, highlighting the advantages and disadvantages of each. The final section concludes with some proposals about the best ways of understnding the complexity that constitutes the Indian cultural reality....

  13. Washington Irving and the American Indian.

    Science.gov (United States)

    Littlefield, Daniel F., Jr.

    1979-01-01

    Some modern scholars feel that Washington Irving vacillated between romanticism and realism in his literary treatment of the American Indian. However, a study of all his works dealing with Indians, placed in context with his non-Indian works, reveals that his attitude towards Indians was intelligent and enlightened for his time. (CM)

  14. Equality in Education for Indian Women.

    Science.gov (United States)

    Krepps, Ethel

    1980-01-01

    Historically, Indian women have been denied education due to: early marriage and family responsibilities; lack of money; inadequate family attention to education; the threat education poses to Indian men; and geographical location. Indian tribes can best administer funds and programs to provide the education so necessary for Indian women. (SB)

  15. PHWR Fuel - an integrated approach in Indian context

    Energy Technology Data Exchange (ETDEWEB)

    Jayaraj, R.N. [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderabad (India)

    2008-07-01

    The nuclear power programme in India is based on a three-stage approach in which the Pressurized Heavy Water Reactors (PHWR) forms the backbone of the first stage. Over the years, apart from gaining expertise in design, construction and operation of PHWRs, innovative fuel designs and manufacturing technologies have also been evolved. Presently, thirteen PHWR 220 units and two PHWR 540 units are in operation. Three more PHWR 220 units are in the advanced stage of construction. In addition, the PHWR power generation programme envisages construction of eight more PHWR 700 units. Nuclear Fuel Complex (NFC) at Hyderabad, established in early 70s, is the only manufacturer of fuel and reactor core structurals for all the PHWRs in India. Since inception, the thrust has been on indigenous development of technology in the areas of production processes, equipment manufacture and quality assurance programmes. Commensurate with the PHWR programme, NFC has expanded its production capacities and has fabricated more than 380,000 fuel bundles since inception. Towards optimization of uranium resources and implementation of 'closed fuel cycle' concept, large quantities of reprocessed uranium fuel bundles have been manufactured and introduced in the initial cores of PHWRs. In recent times, NFC introduced several modifications in the production processes like vapour ammonia precipitation for UO{sub 2} powder production, advanced resistance welding controls and improved versions of welding machines, which all have facilitated in improving the quality and productivity of the fuel. Superior quality control systems like spectrophotometric determination of SSA of UO{sub 2} powders, machine vision systems for pellet inspection, thermography for evaluating weld integrity, etc. has channelised NDT techniques into fuel production lines. The paper summarizes various improvements carried out in the design and manufacture of PHWR fuel. New concepts evolved in high burn-up fuels and

  16. PHWR Fuel - an integrated approach in Indian context

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2008-01-01

    The nuclear power programme in India is based on a three-stage approach in which the Pressurized Heavy Water Reactors (PHWR) forms the backbone of the first stage. Over the years, apart from gaining expertise in design, construction and operation of PHWRs, innovative fuel designs and manufacturing technologies have also been evolved. Presently, thirteen PHWR 220 units and two PHWR 540 units are in operation. Three more PHWR 220 units are in the advanced stage of construction. In addition, the PHWR power generation programme envisages construction of eight more PHWR 700 units. Nuclear Fuel Complex (NFC) at Hyderabad, established in early 70s, is the only manufacturer of fuel and reactor core structurals for all the PHWRs in India. Since inception, the thrust has been on indigenous development of technology in the areas of production processes, equipment manufacture and quality assurance programmes. Commensurate with the PHWR programme, NFC has expanded its production capacities and has fabricated more than 380,000 fuel bundles since inception. Towards optimization of uranium resources and implementation of 'closed fuel cycle' concept, large quantities of reprocessed uranium fuel bundles have been manufactured and introduced in the initial cores of PHWRs. In recent times, NFC introduced several modifications in the production processes like vapour ammonia precipitation for UO 2 powder production, advanced resistance welding controls and improved versions of welding machines, which all have facilitated in improving the quality and productivity of the fuel. Superior quality control systems like spectrophotometric determination of SSA of UO 2 powders, machine vision systems for pellet inspection, thermography for evaluating weld integrity, etc. has channelised NDT techniques into fuel production lines. The paper summarizes various improvements carried out in the design and manufacture of PHWR fuel. New concepts evolved in high burn-up fuels and development of state

  17. The Indian ultrasound paradox

    OpenAIRE

    Akbulut-Yuksel, Mevlude; Rosenblum, Daniel

    2012-01-01

    The liberalization of the Indian economy in the 1990s made prenatal ultrasound technology affordable and available to a large fraction of the population. As a result, ultrasound use amongst pregnant women rose dramatically in many parts of India. This paper provides evidence on the consequences of the expansion of prenatal ultrasound use on sex-selection. We exploit state-by-cohort variation in ultrasound use in India as a unique quasi-experiment. We find that sex-selective abortion of female...

  18. Indian advanced nuclear reactors

    International Nuclear Information System (INIS)

    Saha, D.; Sinha, R.K.

    2005-01-01

    For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

  19. Generation of data base for on-line fatigue life monitoring of Indian nuclear power plant components: Part I - Generation of Green's functions for end fitting

    International Nuclear Information System (INIS)

    Mukhopadhyay, N.K.; Dutta, B.K.; Kushwaha, H.S.

    1994-01-01

    Green's function technique is the heart of the on- line fatigue monitoring methodology. The plant transients are converted to stress and temperature response using this technique. To implement this methodology in a nuclear power plant, Green's functions are to be generated in advance. For structures of complex geometries, Green's functions are to be stored in a data base to convert on-line, the plant data to temperature/stress response, using a personal computer. End fitting, end shield, pressurizer, steam generator tube sheet are few such components of PHWR where fatigue monitoring is needed. In the present paper, Green's functions are generated for end fitting of a 235 MWe Indian PHWR using finite element method. End fitting has been analysed using both 3-D and 2-D (axisymmetric) finite element models. Temperature and stress Green's functions are generated at few critical locations using the code ABAQUS. (author). 10 refs., 11 figs

  20. Indian Vacuum Society: The Indian Vacuum Society

    Science.gov (United States)

    Saha, T. K.

    2008-03-01

    The Indian Vacuum Society (IVS) was established in 1970. It has over 800 members including many from Industry and R & D Institutions spread throughout India. The society has an active chapter at Kolkata. The society was formed with the main aim to promote, encourage and develop the growth of Vacuum Science, Techniques and Applications in India. In order to achieve this aim it has conducted a number of short term courses at graduate and technician levels on vacuum science and technology on topics ranging from low vacuum to ultrahigh vacuum So far it has conducted 39 such courses at different parts of the country and imparted training to more than 1200 persons in the field. Some of these courses were in-plant training courses conducted on the premises of the establishment and designed to take care of the special needs of the establishment. IVS also regularly conducts national and international seminars and symposia on vacuum science and technology with special emphasis on some theme related to applications of vacuum. A large number of delegates from all over India take part in the deliberations of such seminars and symposia and present their work. IVS also arranges technical visits to different industries and research institutes. The society also helped in the UNESCO sponsored post-graduate level courses in vacuum science, technology and applications conducted by Mumbai University. The society has also designed a certificate and diploma course for graduate level students studying vacuum science and technology and has submitted a syllabus to the academic council of the University of Mumbai for their approval, we hope that some colleges affiliated to the university will start this course from the coming academic year. IVS extended its support in standardizing many of the vacuum instruments and played a vital role in helping to set up a Regional Testing Centre along with BARC. As part of the development of vacuum education, the society arranges the participation of

  1. Rasam Indian Restaurant Menu 2017

    OpenAIRE

    Rasam Indian Restaurant

    2017-01-01

    A little bit about us, we opened our doors for business in November 2003 with the solid ambition to serve high quality authentic Indian cuisine in Dublin. Indian food over time has escaped the European misunderstanding or notion of ‘one sauce fits all’ and has been recognised for the rich dining experience with all the wonderful potent flavours of India Rasam wanted to contribute to the Indian food awakening and so when a suitable premise came available in Glasthule at the heart of a busy...

  2. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Author Affiliations. SATYAM MUKHERJEE1. Department of Operations Management, Quantitative Methods & Information Systems; Indian Institute of Management, Udaipur; and Research Center for Open Digital Innovation, Purdue University, IN 47906, USA ...

  3. Indian Academy of Sciences Conference Series | Indian Academy of ...

    Indian Academy of Sciences (India)

    Home; Journals; Indian Academy of Sciences Conference Series; Volume 1; Issue 1. Chimera-like states generated by large perturbation of synchronous state of coupled metronomes. SERGEY BREZETSKIY DAWID DUDKOWSKI PATRYCJA JAROS JERZY WOJEWODA KRZYSZTOF CZOLCZYNSKI YURI MAISTRENKO ...

  4. Co-firing Bosnian coals with woody biomass: Experimental studies on a laboratory-scale furnace and 110 MWe power unit

    Directory of Open Access Journals (Sweden)

    Smajevic Izet

    2012-01-01

    Full Text Available This paper presents the findings of research into cofiring two Bosnian cola types, brown coal and lignite, with woody biomass, in this case spruce sawdust. The aim of the research was to find the optimal blend of coal and sawdust that may be substituted for 100% coal in large coal-fired power stations in Bosnia and Herzegovina. Two groups of experimental tests were performed in this study: laboratory testing of co-firing and trial runs on a large-scale plant based on the laboratory research results. A laboratory experiment was carried out in an electrically heated and entrained pulverized-fuel flow furnace. Coal-sawdust blends of 93:7% by weight and 80:20% by weight were tested. Co-firing trials were conducted over a range of the following process variables: process temperature, excess air ratio and air distribution. Neither of the two coal-sawdust blends used produced any significant ash-related problems provided the blend volume was 7% by weight sawdust and the process temperature did not exceed 1250ºC. It was observed that in addition to the nitrogen content in the co-fired blend, the volatile content and particle size distribution of the mixture also influenced the level of NOx emissions. The brown coal-sawdust blend generated a further reduction of SO2 due to the higher sulphur capture rate than for coal alone. Based on and following the laboratory research findings, a trial run was carried out in a large-scale utility - the Kakanj power station, Unit 5 (110 MWe, using two mixtures; one in which 5%/wt and one in which 7%/wt of brown coal was replaced with sawdust. Compared to a reference firing process with 100% coal, these co-firing trials produced a more intensive redistribution of the alkaline components in the slag in the melting chamber, with a consequential beneficial effect on the deposition of ash on the superheater surfaces of the boiler. The outcome of the tests confirms the feasibility of using 7%wt of sawdust in combination

  5. Application of MELCOR Code to a French PWR 900 MWe Severe Accident Sequence and Evaluation of Models Performance Focusing on In-Vessel Thermal Hydraulic Results

    International Nuclear Information System (INIS)

    De Rosa, Felice

    2006-01-01

    In the ambit of the Severe Accident Network of Excellence Project (SARNET), funded by the European Union, 6. FISA (Fission Safety) Programme, one of the main tasks is the development and validation of the European Accident Source Term Evaluation Code (ASTEC Code). One of the reference codes used to compare ASTEC results, coming from experimental and Reactor Plant applications, is MELCOR. ENEA is a SARNET member and also an ASTEC and MELCOR user. During the first 18 months of this project, we performed a series of MELCOR and ASTEC calculations referring to a French PWR 900 MWe and to the accident sequence of 'Loss of Steam Generator (SG) Feedwater' (known as H2 sequence in the French classification). H2 is an accident sequence substantially equivalent to a Station Blackout scenario, like a TMLB accident, with the only difference that in H2 sequence the scram is forced to occur with a delay of 28 seconds. The main events during the accident sequence are a loss of normal and auxiliary SG feedwater (0 s), followed by a scram when the water level in SG is equal or less than 0.7 m (after 28 seconds). There is also a main coolant pumps trip when ΔTsat < 10 deg. C, a total opening of the three relief valves when Tric (core maximal outlet temperature) is above 603 K (330 deg. C) and accumulators isolation when primary pressure goes below 1.5 MPa (15 bar). Among many other points, it is worth noting that this was the first time that a MELCOR 1.8.5 input deck was available for a French PWR 900. The main ENEA effort in this period was devoted to prepare the MELCOR input deck using the code version v.1.8.5 (build QZ Oct 2000 with the latest patch 185003 Oct 2001). The input deck, completely new, was prepared taking into account structure, data and same conditions as those found inside ASTEC input decks. The main goal of the work presented in this paper is to put in evidence where and when MELCOR provides good enough results and why, in some cases mainly referring to its

  6. Zoogeography of the Indian Ocean

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, T.S.S.

    The distribution pattern of zooplankton in the Indian Ocean is briefly reviewed on a within and between ocean patterns and is limited to species within a quite restricted sort of groups namely, Copepoda, Chaetognatha, Pteropoda and Euphausiacea...

  7. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Department of Aerospace Engineering, Indian Institute of Science, Bangalore 560012, India; Structures group, ISRO Satellite Centre, Bangalore 560017, India; Department of Mechanical Engineering, PES University, Bangalore 560085, India ...

  8. Oceanography of the Indian Ocean

    Digital Repository Service at National Institute of Oceanography (India)

    Desai, B.N.

    This volume is an outcome of the presentation of selected 74 papers at the International Symposium on the Oceanography of the Indian Ocean held at National Institute of Oceanography during January 1991. The unique physical setting of the northern...

  9. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    dependent Phase Stability, TEM Address: Dept. of Materials Engineering, Indian Institute of Science, Bengaluru 560 012, Karnataka Contact: Office: (080) 2293 2834. Residence: 99006 26327. Email: csrivastava@materials.iisc.ernet.in. YouTube ...

  10. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Srinivasa Raghavan, Dr N R . Date of birth: 28 May 1972. Specialization: Decision Sciences & Technologies Address during Associateship: Department of Maagement Studies, Indian Institute of Science, Bengaluru 560 012. YouTube; Twitter ...

  11. Indian Institute of Technology, Guwahat

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Refresher Course in Experimental Physics – Indian Institute of Technology, ... Information and Announcements Volume 10 Issue 1 January 2005 pp 96-96 ...

  12. Home | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    2016-08-24

    Aug 24, 2016 ... Ayurveda, the Indian traditional medical system, on the other hand, has always ... as a holistic response of an individual to the environmental challenge. ... has been effective in the translation of network medicine into clinical ...

  13. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    TCP performs poorly in wireless mobile networks due to large bit error rates. ... TCP, and find considerable improvement in data throughput over wireless links. ... Centre for Electronics Design and Technology, Indian Institute of Science, ...

  14. Polydactyly in the American Indian.

    Science.gov (United States)

    Bingle, G J; Niswander, J D

    1975-01-01

    Polydactyly has an incidence in the American Indian twice that of Caucasians. A minimum estimate of this incidence is 2.40 per 1,000 live births. Preaxial type 1 has an incidence three to four times that reported for Caucasians or Negroes. The overall sex ratio in Indians is distorted with more males affected than females. The preaxial type 1 anomaly has a strong predilection for the hands and always is unilateral in contrast to postaxial type B where more than one-half are bilateral. The evidence to date, consisting of varying incidences of specific types of polydactyly among American whites, Negroes, and Indians in varying enviroments, suggests different gene-frequencies for polydactyly in each population. The incidence in Indians with 50% Caucasian admixture suggests that the factors controlling polydactyly are in large part genetically determined. Family studies and twin studies reported elsewhere offer no clear-cut genetic model which explains the highly variable gene frequencies.

  15. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Department of Pharmacology, Institute of PG Medical Education ... Address: Department of Chemistry, Indian Institute of Technology, Kharagpur 721 302, W.B.. Contact: ... Specialization: Elementary Particle Physics, Field Theory and ...

  16. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    ... their information technology (IT) related activities to third party software companies. Indian software companies have become leaders in providing these services. Companies from several other countries are also competing for the top slot.

  17. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    .D. (Bangalore), FNASc. Date of birth: 4 May 1968. Specialization: Astrosat Mission & UV Studies, Stellar Population, Nearby Galaxies, Star Clusters, Stellar Evolution, Galactic Dynamics Address: Indian Institute of Astrophysics, Sarjapur Road, ...

  18. Environmental Protection in Indian Country

    Science.gov (United States)

    EPA's efforts to protect human health and the environment of federally recognized Indian tribes by supporting implementation of federal environmental laws consistent with the federal trust responsibility, and the government-to-government relationship.

  19. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Associate Profile. Period: 2001–2005. Satheesh, Dr S K . Date of birth: 1 May 1970. Specialization: Aerosols in Climate Address during Associateship: Centre for Atmospheric & Oceanic, Sciences, Indian Institute of Science, Bangalore 560 012

  20. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Anand, Dr V G . Specialization: Bio-inorganic Chemistry, Pi-Conjugated Macrocycles, Supramolecular Chemistry Address during Associateship: Indian Institute of Science Edn., and Research, 900, NCL Innovation Park, Pashan, Pune 411 008

  1. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Last known address: Department of Mathematics, Purdue University, West Lafayette, Indiana 47907, USA. Elected: .... Last known address: Professor, Department of Physics, Indian Institute of Science, Bengaluru 560 012 ...... Madhu Sudan

  2. Development of Indian passenger transport

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, R. [Indira Ghandi Institute of Development Research, Mumbai (India)

    1998-05-01

    The Indian transport sector has been studied using logistic substitution. The share of rail transport is declining, while road and air transport are increasing. These developments are not desirable from an energy-efficiency perspective. (author)

  3. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Associate Profile. Period: 1993–1996. Das, Dr P P . Date of birth: 30 July 1961. Specialization: Computer Engineering Address during Associateship: Dept. of Computer Science and, Engineering, Indian Institute of Technology, Kharagpur 721 302.

  4. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Home; Fellowship; Associateship. Associate Profile. Period: 1983–1986. Guru Row, Dr T N . Date of birth: 26 September 1951. Specialization: Crystallography Address during Associateship: Solid State and Structural, Chemistry Unit, Indian ...

  5. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Associate Profile. Period: 1983–1986. Krishnamurthy, Prof. H R . Date of birth: 21 September 1951. Specialization: Theory of Magnetism Address during Associateship: Department of Physics, Indian Institute of Science, Bengaluru 560 012.

  6. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Period: 1990–1994. Patel, Dr A D . Date of birth: 17 January 1959. Specialization: Particle Theory Address during Associateship: Centre for Theoretical Studies, Indian Institute of Science, Bangalore 560 012. YouTube; Twitter; Facebook; Blog ...

  7. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Checkpointing is the process of saving the status information. ... Supercomputer Education and Research Centre (SERC), Indian Institute of Science, Bangalore 560 ... Manuscript received: 27 August 1998; Manuscript revised: 8 June 2000 ...

  8. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    ... VLSI clock interconnects; delay variability; PDF; process variation; Gaussian random ... Supercomputer Education and Research Centre, Indian Institute of Science, ... Manuscript received: 27 February 2009; Manuscript revised: 9 February ...

  9. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address during Associateship: Non-Ferrous Process Division, National ... A revised version of the document 'Scientific Values: Ethical Guidelines and ... 4 to 6 November 2016 at Indian Institute of Science Education and Research, Bhopal.

  10. Home | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    2016-12-23

    Dec 23, 2016 ... ... hosted by the Indian Institute of Science Education and Research, ... that draws upon several different areas of modern mathematics such as ... He spoke of his experiences in Rajasthan, where, by use of traditional methods, ...

  11. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Address: Centre for Biomedical Engineering, Indian Institute of Technology, ..... Bag, Dr Amulya Kumar ..... Specialization: Atmospheric Sciences, Global Change & Atmospheric Environment, Urban Air Pollution & Chemical-Climate Change, ...

  12. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Duke). Date of birth: 24 May 1962. Specialization: Algorithms (Sequential & Parallel), Probabilistic Analysis & Randomization and Computational Geometry Address: Department of Computer Science & Engineering, Indian Institute of Technology, ...

  13. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Date of birth: 1 July 1959. Specialization: Game Theory & Mechanism Design, Electronic Commerce Internet and Network Economics Address: Department of Computer Science & Automation, Indian Institute of Science, Bengaluru 560 012, Karnataka Contact: Office: (080) 2293 2773. Residence: (080) 2331 0265

  14. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    , Dr Manindra. Date of birth: 20 May 1966. Specialization: Computer Science and Engineering Address during Associateship: Dept. of Computer Science & Engg., Indian Institute of Technology, Kanpur 208 016. YouTube; Twitter; Facebook ...

  15. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Specialization: Databases, Real-Time Systems, Use of Information & Communication Technology for Socioeconomic Development Address: Department of Computer Science & Engineering, Indian Institute of Technology, Powai, Mumbai 400 076, Maharashtra Contact: Office: (022) 2576 7740. Residence: (022) 2576 8740

  16. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    .D. (UC, Berkeley). Date of birth: 14 April 1969. Specialization: Web Search & Mining, Graph Information Retrieval Address: Department of Computer Science & Engineering, Indian Institute of Technology, Powai, Mumbai 400 076, Maharashtra

  17. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Specialization: Computer Science & Engineering, Information Technology and Electronics Address: INSA Senior Scientist, Faculty Consciousness Studies Programme, National Institute of Advanced Studies, Indian Institute of Science Campus, Bengaluru 560 012, Karnataka Contact: Residence: (080) 2360 2635

  18. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Author Affiliations. NEENA ISAAC1 2 T I ELDHO1. Department of Civil Engineering, Indian Institute of Technology Bombay, Mumbai 400076, India; Central Water and Power Research Station, Khadakwasla, Pune 411024, India ...

  19. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Author Affiliations. TAPAS KARMAKER1 RANJAN DAS2. Department of Civil Engineering, Thapar University, Patiala 147004, India; School of Mechanical, Materials, and Energy Engineering, Indian Institute of Technology Ropar, Rupnagar 140001, India ...

  20. Sadhana | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Radar-based hydrological studies in various countries have proven that ... for hydrological modelling and/or flood-related studies in Indian river basins. ... in the runoff volume was small, but the difference in the peak flow was substantial.

  1. Taxation and the American Indian

    Science.gov (United States)

    Brunt, David

    1973-01-01

    The article explores American Indian tribal rights to tax exemptions and self-imposed taxation; general recommendations on possible tribal tax alternatives; and evaluation of the probable economic effect of taxation. (FF)

  2. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Date of birth: 6 January 1981 ... Date of birth: 19 February 1985 .... Address: School of Basic Sciences, Indian Institute of Technology, Mandi 175 005, H.P. ... Specialization: Game Theory & Optimisation, Stochastic Control, Information Theory

  3. Fellowship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Mobile: 94797 25236 ... Address: Managing Director, Techcellence Consultancy Services, Pvt. Ltd., 5, Pushkaraj, Pushpak .... Address: Department of Computer Science & Automation, Indian Institute of Science, .... http://nayak.web.cern.ch.

  4. Associateship | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Period: 1994–1998. Rangarajan, Dr P N . Date of birth: 15 April 1963. Specialization: Biochemistry Address during Associateship: Department of Biochemistry, Indian Institute of Science, Bengaluru 560 012. YouTube; Twitter; Facebook; Blog ...

  5. Internationalization Of Indian IT Multinationals

    OpenAIRE

    Singh, Abhishek

    2009-01-01

    Indian IT industry has emerged to be a strong and influential player on the world map. The industry which did not existed a few decades ago is now a major exporter of software services to major markets. The Indian IT firms now seem to move beyond exporting and advance further into the international market. With the help of case study approach, this study tends to examine the internationalization of these firms. The dissertation is aimed to see how far the traditional theories o...

  6. A case study for INPRO methodology based on Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    Anantharaman, K.; Saha, D.; Sinha, R.K.

    2004-01-01

    Under Phase 1A of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) a methodology (INPRO methodology) has been developed which can be used to evaluate a given energy system or a component of such a system on a national and/or global basis. The INPRO study can be used for assessing the potential of the innovative reactor in terms of economics, sustainability and environment, safety, waste management, proliferation resistance and cross cutting issues. India, a participant in INPRO program, is engaged in a case study applying INPRO methodology based on Advanced Heavy Water Reactor (AHWR). AHWR is a 300 MWe, boiling light water cooled, heavy water moderated and vertical pressure tube type reactor. Thorium utilization is very essential for Indian nuclear power program considering the indigenous resource availability. The AHWR is designed to produce most of its power from thorium, aided by a small input of plutonium-based fuel. The features of AHWR are described in the paper. The case study covers the fuel cycle, to be followed in the near future, for AHWR. The paper deals with initial observations of the case study with regard to fuel cycle issues. (authors)

  7. 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

  8. NAPP liquid shutoff rod system design, development, testing and precommissioning feed back study

    International Nuclear Information System (INIS)

    Soni, K.L.; Kaushik, R.V.; Mahajan, S.V.

    1989-01-01

    The development testing of a liquid poison shutoff rod system has enabled the evolution of a proven and acceptable design of the secondary shutdown system for 235 MWe standardised Pressurised Heavy Water Reactors (PHWRs). The availability of a full scale test loop is also proving for checking and qualifying the various suggestions for online improvements in the system. (author)

  9. Leading Indian Business-Groups

    Directory of Open Access Journals (Sweden)

    Maria Alexandrovna Vorobyeva

    2016-01-01

    Full Text Available The goal of this paper is to investigate the evolution of the leading Indian business-groups under the conditions of economical liberalization. It is shown that the role of modern business-groups in the Indian economy is determined by their high rate in the gross domestic product (GDP, huge overall actives, substantial pert in the e[port of goods and services, as well as by their activities in modern branch structure formatting, and developing labor-intensive and high-tech branches. They strongly influence upon economical national strategies, they became a locomotive of internationalization and of transnationalization of India, the basis of the external economy factor system, the promoters of Indian "economical miracle" on the world scene, and the dynamical segment of economical and social development of modern India. The tendencies of the development of the leading Indian business groups are: gradual concentration of production in few clue sectors, "horizontal" structure, incorporation of the enterprises into joint-stock structure, attraction of hired top-managers and transnationaliziation. But against this background the leading Indian business-groups keep main traditional peculiarities: they mostly still belong to the families of their founders, even today they observe caste or communal relations which are the basis of their non-formal backbone tides, they still remain highly diversificated structures with weak interrelations. Specific national ambivalence and combination of traditions and innovations of the leading Indian business-groups provide their high vitality and stability in the controversial, multiform, overloaded with caste and confessional remains Indian reality. We conclude that in contrast to the dominant opinion transformation of these groups into multisectoral corporations of the western type is far from completion, and in the nearest perspective they will still possess all their peculiarities and incident social and economical

  10. New associates | Announcements | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Translational Health Science and Technology Institute, Faridabad. Praveen Kumar Indian Institute of Science, Bengaluru. S Mishra Sabyashachi Mishra Indian Institute of Technology, Kharagpur. Jagannath Mondal TIFR Centre for Interdisciplinary Sciences, Hyderabad. Samrat Mondal Wildlife Institute of India, Dehradun.

  11. Journal of Genetics | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    GJB2 and GJB6 gene mutations found in Indian probands with congenital hearing impairment .... and plasma factor VII coagulant activity in Asian Indian families predisposed to .... Tetrasomy 18p in a male dysmorphic child in southeast Turkey.

  12. Superficial mineral resources of the Indian Ocean

    Digital Repository Service at National Institute of Oceanography (India)

    Siddiquie, H.N.; Hashimi, N.H.; Gujar, A; Valsangkar, A

    The sea floor of the Indian Ocean and the continental margins bordering the ocean are covered by a wide variety of terrigenous, biogenous and anthigenic mineral deposits. The biogenous deposits in the Indian Ocean comprise the corals on shallow...

  13. Journal of Genetics | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    ... Sharat Chandra (both of Indian Institute of Science, Bengaluru) and Suresh Jayakar ... In 1985, the Indian Academy of Sciences, Bengaluru, revived publication of ... It publishes papers and review articles on current topics, commentaries and ...

  14. Utilizing linkage disequilibrium information from Indian Genome ...

    Indian Academy of Sciences (India)

    Using LD information derived from Indian Genome Variation database (IGVdb) on populations .... Line diagram represents the SNPs selected in Indian (upper panel) and CEPH .... out procedure for extracting DNA from human nucleated cells.

  15. The Comprehensive View of Indian Education.

    Science.gov (United States)

    Kaegi, Gerda

    Relating historical conflicts between Indians and whites, the document explained how education was originally aimed at "civilizing" and domesticating the Canadian Indian. This philosophy, used extensively by church groups that established the original Indian schools, alienated children from both the white society and the educational…

  16. New fellows | Announcements | Indian Academy of Sciences

    Indian Academy of Sciences (India)

    Aninda J Bhattacharyya, Indian Institute of Science, Bengaluru; Suvendra N Bhattacharyya, CSIR-Indian Institute of Chemical Biology, Kolkata; Mitali Chatterjee, Institute of Postgraduate Medical Education & Research, Kolkata; Prasanta K Das, Indian Association for the Cultivation of Science, Kolkata; Swapan K Datta, ...

  17. History and Acculturation of the Dakota Indians.

    Science.gov (United States)

    Satterlee, James L.; Malan, Vernon D.

    Relating the history of the Dakota Indians from their origins to the present time, this document also examines the effects of acculturation on these Sioux people. Beginning with the Paleo-Indians of North America, it details the structure of the Dakota culture and attempts to acculturate the Indians into white society. Historical and current…

  18. U. S. and Canadian Indian Periodicals.

    Science.gov (United States)

    Price, John

    The document lists and discusses Indian-published and Indian-oriented newspapers, periodicals, and other assorted publications generally designed to establish a communication system reflecting the interest of the majority of American Indians. Also provided are resumes of several publications that are thought to have gained wide acceptance through…

  19. Promoting Indian Library Use. Guide Number 7.

    Science.gov (United States)

    Townley, Charles T.

    Individuals, organizations, and American Indian tribes are rapidly recognizing the value of libraries. They are recognizing that libraries and the information services which they offer are necessary to meet Indian goals. Specific sensitivity to Indian ways and alternatives is just developing as library and information services develop in Indian…

  20. Congressional Social Darwinism and the American Indian

    Science.gov (United States)

    Blinderman, Abraham

    1978-01-01

    Summarizing a congressional report on civil and military treatment of American Indians, this article asserts that the social Darwinism of the day prevailed among all congressional committee members ("Even friends of the Indian... knew American expansionism, technology, and racial ideology would reduce the Indian to a pitiful remnant...) (JC)