WorldWideScience

Sample records for reactor operator performance

  1. Performance Indicators of Operating Reactors

    Data.gov (United States)

    Nuclear Regulatory Commission — A list of Performance Indicators (PI) that are reported to the NRC by licensees at the end of each quarter in accordance with Inspection Manual Chapters (IMC) 0608,...

  2. Research about reactor operator's personality characteristics and performance

    International Nuclear Information System (INIS)

    Wei Li; He Xuhong; Zhao Bingquan

    2003-01-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  3. Operational limitations of light water reactors relating to fuel performance

    International Nuclear Information System (INIS)

    Cheng, H.S.

    1976-07-01

    General aspects of fuel performance for typical Boiling and Pressurized Water Reactors are presented. Emphasis is placed on fuel failures in order to make clear important operational limitations. A discussion of fuel element designs is first given to provide the background information for the subsequent discussion of several fuel failure modes that have been identified. Fuel failure experiences through December 31, 1974, are summarized. The operational limitations that are required to mitigate the effects of fuel failures are discussed

  4. Overcoming the effects of stress on reactor operator performance

    International Nuclear Information System (INIS)

    He Xuhong; Wei Li; Zhao Bingquan

    2003-01-01

    Reactor operators may be exposed to significant levels of stress during plant emergencies and their performance may be affected by the stress. This paper first identified the potential sources of stress in the nuclear power plant, then discussed the ways in which stress is likely to affect the reactor operators, and finally identified several training approaches for reducing or eliminating stress effects. The challenges for effective stress reducing training may seem daunting, yet the challenges are real and must be addressed. This paper reviewed researches in training design, knowledge and skill acquisition, and training transfer point to a number of strategies that can be used to address these challenges and lead to more effective training and development. (author)

  5. Overcoming the effects of stress on reactor operator performance

    Energy Technology Data Exchange (ETDEWEB)

    He Xuhong; Wei Li; Zhao Bingquan [Tsinghua Univ., Nuclear Power Plant Simulation Training Center, Beijing (China)

    2003-03-01

    Reactor operators may be exposed to significant levels of stress during plant emergencies and their performance may be affected by the stress. This paper first identified the potential sources of stress in the nuclear power plant, then discussed the ways in which stress is likely to affect the reactor operators, and finally identified several training approaches for reducing or eliminating stress effects. The challenges for effective stress reducing training may seem daunting, yet the challenges are real and must be addressed. This paper reviewed researches in training design, knowledge and skill acquisition, and training transfer point to a number of strategies that can be used to address these challenges and lead to more effective training and development. (author)

  6. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  7. Value addition initiatives for CANDU reactor operation performance

    International Nuclear Information System (INIS)

    Chugh, V.; Parmar, R.; Schut, J.; Sherin, J.; Xie, H.; Zobin, D.

    2013-01-01

    Recently, AMEC NSS initiated projects for CANDU® station performance engineering with potentially high returns for the utilities. This paper discusses three initiatives. Firstly, optimization of instrument calibration interval from 1 to 3 years will reduce time commitments on the maintenance resources on top of financial savings ~$3,500 per instrument. Secondly, reactor thermal power uncertainty assessment shows the level of operation which is believed to have an over-conservative margin that can be used to increase power by up to 0.75%. Finally, as an alternative means for controlling Reactor Inlet Header Temperature (RIHT), physical modifications to the High Pressure (HP) feedwater heaters can be useful for partially recovering RIHT resulting in increased production by 10-12 MWe. (author)

  8. Operating performance of the prototype heavy water reactor Fugen

    International Nuclear Information System (INIS)

    1984-01-01

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

  9. Fort St. Vrain reactor performance and operation to full power

    International Nuclear Information System (INIS)

    Simon, W.A.; Bramblett, G.C.

    1982-01-01

    The Fort St. Vrain Nuclear Generating Station, powered by a high-temperature gas-cooled reactor (HTGR), has now been tested to full thermal power. Testing was conducted for the dual purposes of demonstrating component and system capability as a part of the rise-to-power program and determining core fluctuation/redistribution behavior under full power conditions. Both objectives were met. Full power performance of all major components and the achievement of nearly all design objectives has been verified. In addition, the tests showed that the fluctuation phenomenon has been corrected. Core region outlet temperature redistributions have been characterized, related to a physical mechanism, and shown to be inconsequential for overall plant operation

  10. Research about reactor operator's personability characteristics and performance

    Energy Technology Data Exchange (ETDEWEB)

    Wei Li; He Xuhong; Zhao Bingquan [Tsinghua Univ., Institute of Nuclear Energy Technology, Beijing (China)

    2003-03-01

    To predict and evaluate the reactor operator's performance by personality characteristics is an important part of reactor operator safety assessment. Using related psychological theory combined with the Chinese operator's fact and considering the effect of environmental factors to personality analysis, paper does the research about the about the relationships between reactor operator's performance and personality characteristics, and offers the reference for operator's selection, using and performance in the future. (author)

  11. PLUS7TM In-Reactor Operating Performance and Economics

    International Nuclear Information System (INIS)

    Kim, Kyutae; Jang, Youngki; Choi, Joonhyung; Lee, Jinseok; Kim, Yoonho; Suh, Jungmin

    2006-01-01

    KNFC has developed an advanced fuel, PLUS7 TM , for the Korean Standard Nuclear Power Plants(KSNPs) through the joint development program with Westinghouse. With the help of various out-of-pile tests, it is found that the PLUS7 TM shows much better performance than the current fuel, GUARDIAN TM from the safety and economy points of view. Now four Lead Test Assembles(LTAs) of the PLUS7 TM are being irradiated for the 3 rd cycle after the successful completion of the 1 st and 2 nd irradiation cycles. During the 1 st and 2 nd irradiation cycles, no fuel failure was observed at LTAs and their nuclear-related parameters matched their design values well. During the overhaul period, on the other hand, pool side examinations were performed for four LTAs to generate key in-reactor fuel performance data such as fuel rod and assembly growths, fuel rod-to-top nozzle gap, fuel assembly bow and twist, fuel rod bow, spacer grid width, fuel rod diameter and fuel rod oxide layer thickness. It is found that all measured values are bounded by upper and lower predicted ones. The detailed economic analyses have shown that significant fuel cycle cost can be reduced by more than one million dollars per cycle of one KSNP with the introduction of the PLUS7 TM assembly. Furthermore, more than one hundred million dollars with power up-rating of 5% can be saved annually for currently operating eight KSNPs, which is easily and safety achievable with the PLUS7 TM assembly

  12. Regulations for RA reactor operation

    International Nuclear Information System (INIS)

    1980-09-01

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions [sr

  13. Operation and Performance of the Supercritical Fluids Reactor (SFR)

    National Research Council Canada - National Science Library

    Hanush, R

    1996-01-01

    The Supercritical Fluids Reactor (SFR) at Sandia National Laboratories, CA has been developed to examine and solve engineering, process, and fundamental chemistry issues regarding the development of supercritical water oxidation (SCWO...

  14. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  15. Reactor operational transient analysis

    International Nuclear Information System (INIS)

    Shin, W.K.; Chae, S.K.; Han, K.I.; Yang, K.S.; Chung, H. D.; Kim, H.G.; Moon, H.J.; Ryu, Y.H.

    1983-01-01

    To build up efficient capability of safety review and inspection for the nuclear power plants, four area of studies have performed as follows: 1) In order to search the most optimized operating method during load follow operating schemes, automatic control and normal control, are compared each other under the CAOC condition. The analysis performed by DDID code has shown that the reactor has to be controlled by the operator manually during load follow operation. 2) Through the sensitivity analysis by COBRA code, the operating parameters, such as coolant pressure, flow rate, inlet temperature, and power distribution are shown to be important to the determination of DNBR. Expecially, inlet temperature of primary coolant system is appeared as the most senstive parameter on DNBR. 3) FRAPCON code is adapted to study the sensitivity of several operational parameters on the mechanical properties of reactor fuel rod. 4) The calculations procedure which is required to be obtained the neutron fluence at the reactor vessel and the spectrum at the surveillance capsule is established. The results of computation are conpared with those of FSAR and SWRI report and proved its applicability to reactor surveillance program. (Author)

  16. Catalytic performance improvement of styrene hydrogenation in trickle bed reactor by using periodic operation

    International Nuclear Information System (INIS)

    Wongkia, Atittahn; Praserthdam, Piyasan; Assabumrungrat, Suttichai; Suriye, Kongkiat; Nonkhamwong, Anuwat

    2013-01-01

    We investigated the catalytic performance improvement of styrene hydrogenation in a trickle bed reactor by using periodic operation. The effects of cycle period and split on relative conversion, which is defined as styrene conversion obtained from periodic operation over that from steady state operation, were examined at various operating conditions including gas and average liquid flow rates, pressure and temperature. The experimental results reveal that both cycle period and split have strong influence on the catalytic performance. The fast mode (short cycle period) is a favorable condition. The improvement by the periodic operation becomes less pronounced for operations at high average liquid flow rate, pressure and temperature. From this study, a maximum improvement of styrene conversion of 18% is observed

  17. Catalytic performance improvement of styrene hydrogenation in trickle bed reactor by using periodic operation

    Energy Technology Data Exchange (ETDEWEB)

    Wongkia, Atittahn; Praserthdam, Piyasan; Assabumrungrat, Suttichai [Chulalongkorn University, Bangkok (Thailand); Suriye, Kongkiat; Nonkhamwong, Anuwat [SCG Chemicals Co. Ltd., Bangkok (Thailand)

    2013-03-15

    We investigated the catalytic performance improvement of styrene hydrogenation in a trickle bed reactor by using periodic operation. The effects of cycle period and split on relative conversion, which is defined as styrene conversion obtained from periodic operation over that from steady state operation, were examined at various operating conditions including gas and average liquid flow rates, pressure and temperature. The experimental results reveal that both cycle period and split have strong influence on the catalytic performance. The fast mode (short cycle period) is a favorable condition. The improvement by the periodic operation becomes less pronounced for operations at high average liquid flow rate, pressure and temperature. From this study, a maximum improvement of styrene conversion of 18% is observed.

  18. Reactor operation monitor

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1982-01-01

    Purpose: To improve the working performance of a reactor by extending the range for the power conditioning due to the control rod operation and flow rate control. Constitution: The results of calculations for the power distribution and the burn-up degree distribution of the reactor core from a reactor performance computer that processes each of measuring signals in a nuclear power plant are used as the inputs for a computing device of the fuel rod power hysteresis to form the power hysteresis for each of the fuel rods up to the present time. The data are used as the inputs for the computing device of the fuel rod performance index, and the fuel rod performance index representing the critical values for the stresses in the fuel rod cladding tubes and the critical values for the duration of the stresses determined from the power hysteresis and the burn-up degree of the fuel rod are calculated for each of the fuel rods. Accordingly, the power conditioning can be carried out upon power-up in the reactor while monitoring the fuel rod performance index f(t) for each of the fuel assemblies, whereby the range for the power conditioning due to the control rod operation and the flow rate control can be extended relative to fuel assemblies in which f(t) is smaller than 1. (Yoshino, Y.)

  19. Performance of a continuously operated flocculent sludge UASB reactor with slaughterhouse wastewater

    Energy Technology Data Exchange (ETDEWEB)

    Sayed, S.; Zeeuw, W. de

    1988-01-01

    This investigation was carried out to assess the performance of a continuously operated, one-stage, flocculent sludge upflow anaerobic sludge blanket (UASB) reactor treating slaughterhouse wastewater at a process temperature of 30/sup 0/C. The results indicate that the type of substrate ingredients, coarse suspended solids, colloidal and soluble compounds in the wastewater, affect the performance of the reactor because of different mechanisms involved in their removal and their subsequent conversion into methane. Two different mechanisms are distinguished. An entrapment mechanism prevails for the elimination of coarse suspended solids while an adsorption mechanism is involved in the removal of the colloidal and soluble fractions of the wastewater. The results obtained lead to the conclusion that the system can satisfactorily handle organic space loads up to 5 kg COD m/sup -3/ day/sup -1/ at 30/sup 0/C. The data indicate, however, that continuing heavy accumulation of substrate components in the reactor is detrimental to the stability of the anaerobic treatment process as the accumulation can lead to sludge flotation and consequently to a complete loss of the active biomass from the reactor.

  20. Power reactors operational diagnosis

    International Nuclear Information System (INIS)

    Dach, K.; Pecinka, L.

    1976-01-01

    The definition of reactor operational diagnostics is presented and the fundamental trends of research are determined. The possible sources of power reactor malfunctions, the methods of defect detection, the data evaluation and the analysis of the results are discussed in detail. In view of scarcity of a theoretical basis and of insufficient in-core instrumentation, operational diagnostics cannot be as yet incorporated in a computer-aided reactor control system. (author)

  1. Licensed operating reactors

    International Nuclear Information System (INIS)

    1990-04-01

    The Operating Units Status Report --- Licensed Operating Reactors provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff on NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non- power reactors in the US

  2. Improvement of pulsing operation performance in the Nuclear Safety Research Reactor (NSRR)

    International Nuclear Information System (INIS)

    Kobayasi, S.; Ishijima, K.; Tanzawa, S.; Fujishiro, T.; Horiki, O.

    1990-01-01

    The Nuclear Safety Research Reactor (NSRR) is one of the TRIGA-type research reactors widely used in the world, and has mainly been used for studying reactor fuel behaviour during postulated reactivity-initiated accidents (RIAs). Its limited pulsing operation capability, however, could produce only a power burst from low power level simulating an RIA event from essentially zero power level. A computerized automatic reactor control system was developed and installed in the NSRR to simulate a wide range of abnormal events in nuclear power plants. This digitalized reactor control system requires no manipulation of the control rods by reactor operators during the course of the pulsing operation. Using this fully automated operation system, a variety of power transients such as power ramping, power bursts from high power level, and so on were made possible with excellent stability and safety. The present modification work in the NSRR and its fruitful results indicate new possibilities in the utilization of the TRIGA type research reactor

  3. Pressurised water reactor operation

    International Nuclear Information System (INIS)

    Birnie, S.; Lamonby, J.K.

    1987-01-01

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

  4. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  5. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

    This update, which appears regularly in each issue of Nuclear Safety, surveys the operations of those power reactors in the US which have been issued operating licenses. Table 1 shows the number of such reactors and their net capacities as of September 30, 1987, the end of the three-month period covered in this report. Table 2 lists the unit capacity and forced outage rate for each licensed reactor for each of the three months (July, August, and September 1987) covered in this report and the cumulative values of these parameters since the beginning of commercial operation. In addition to the tabular data, this article discusses other significant occurrences and developments that affected licensed US power reactors during this reporting period. Status changes at Braidwood Unit 1, Nine Mile Point 2, and Beaver Valley 2 are discussed. Other occurrences discussed are: retraining of control-room operators at Peach Bottom; a request for 25% power for Shoreham, problems at Fermi 2 which delayed the request to go to 75% power; the results of a safety study of the N Reactor at Hanford; a proposed merger of Pacific Gas and Electric with Sacramento Municipal Utility District which would result in the decommissioning of Rancho Seco; the ordered shutdown of Oyster Creek; a minor radioactivity release caused by a steam generator tube rupture at North Anna 1; and 13 fines levied by the NRC on reactor licensees

  6. Licensed operating reactors

    International Nuclear Information System (INIS)

    1989-08-01

    THE OPERATING UNITS STATUS REPORT - LICENSED OPERATING REACTORS provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management from the Headquarters staff of NRC's Office of Enforcement (OE), from NRC's Regional Offices, and from utilities. The three sections of the report are: monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, OE Headquarters and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor-years of experience and non-power reactors in the US

  7. Reactor operation method

    International Nuclear Information System (INIS)

    Osumi, Katsumi; Miki, Minoru.

    1979-01-01

    Purpose: To prevent stress corrosion cracks by decreasing the dissolved oxygen and hydrogen peroxide concentrations in the coolants within a reactor container upon transient operation such as at the start-up or shutdown of bwr type reactors. Method: After a condensate has been evacuated, deaeration operation is conducted while opening a main steam drain line, as well as a main steam separation valve and a by-pass valve in a turbine by-pass line connecting the main steam line and the condenser without by way of a turbine, and the reactor is started-up by the extraction of control rods after the concentration of dissolved oxygen in the cooling water within a pressure vessel has been decreased below a predetermined value. Nuclear heating is started after the reactor water has been increased to about 150 0 C by pump heating after the end of the deaeration operation for preventing the concentration of hydrogen peroxide and oxygen in the reactor water from temporarily increasing immediately after the start-up. The corrosive atmosphere in the reactor vessel can thus be moderated. (Horiuchi, T.)

  8. FREC-4A: a computer program to predict fuel rod performance under normal reactor operation

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Izumi, Fumio

    1981-10-01

    The program FREC-4A (Fuel Reliability Evaluation Code-version 4A) is used for predicting fuel rod performance in normal reactor operation. The performance is calculated in accordance with the irradiation history of fuel rods. Emphasis is placed on the prediction of the axial elongation of claddings induced by pellet-cladding mechanical interaction, including the influence of initially preloaded springs inserted in fuel rod lower plenums. In the FREC-4A, an fuel rod is divided into axial segments. In each segment, it is assumed that the temperature, stress and strain are axi-symmetrical, and the axial strain in constant in fuel pellets and in a cladding, though the values in the pellets and in the cladding are different. The calculation of the contact load and the clearance along the length of a fuel rod and the stress and strain in each segment is explained. The method adopted in the FREC-4A is simple, and suitable to predict the deformation of fuel rods over their full length. This report is described on the outline of the program, the method of solving the stiffness equations, the calculation models, the input data such as irradiation history, output distribution, material properties and pores, the printing-out of input data and calculated results. (Kako, I.)

  9. Licensed operating reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  10. Licensed operating reactors

    International Nuclear Information System (INIS)

    Hartfield, R.A.

    1990-03-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  11. Licensed operating reactors

    International Nuclear Information System (INIS)

    1989-08-01

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  12. Method of reactor operation

    International Nuclear Information System (INIS)

    Maeda, Katsuji.

    1982-01-01

    Purpose: To prevent stress corrosion cracks in stainless steels caused from hydrogen peroxide in reactor operation in which the density of hydrogen peroxide in the reactor water is controlled upon reactor start-up. Method: A heat exchanger equipped with a heat source for applying external heat is disposed into the recycling system for reactor coolants. Upon reactor start-up, the coolants are heated by the heat exchanger till arriving at a temperature at which the dissolving rate is faster than the forming rate of hydrogen peroxide in the coolants, and nuclear heating is started after reaching the above temperature. The temperature of the reactor water is increased in such a manner and, when it arrives at 140 0 C, extraction of control elements is started and the heat source for the heat exchanger is interrupted simultaneously. In this way spikes in the density of hydrogen peroxide are suppressed upon reactor start-up to thereby decrease the stress corrosion cracks in stainless steels. (Horiuchi, T.)

  13. Small reactor operating mode

    International Nuclear Information System (INIS)

    Snell, V.G.

    1997-01-01

    There is a potential need for small reactors in the future for applications such as district heating, electricity production at remote sites, and desalination. Nuclear power can provide these at low cost and with insignificant pollution. The economies required by the small scale application, and/or the remote location, require a review of the size and location of the operating staff. Current concepts range all the way from reactors which are fully automatic, and need no local attention for days or weeks, to those with reduced local staff. In general the less dependent a reactor is on local human intervention, the greater its dependence on intrinsic safety features such as passive decay heat removal, low-stored energy and limited reactivity speed and depth in the control systems. A case study of the design and licensing of the SLOWPOKE Energy System heating reactor is presented. (author)

  14. Evaluation of the dual digestion system 2: operation and performance of the pure oxygen aerobic reactor

    CSIR Research Space (South Africa)

    Messenger, JR

    1993-07-01

    Full Text Available In a comprehensive study of the performance of a full-scale (45 m3) pure oxygen autothermal thermophilic aerobic reactor of a sewage sludge dual digestion system, it was found that: Biological heat generation rate was directly proportional...

  15. Licensed operating reactors

    International Nuclear Information System (INIS)

    1989-11-01

    The US Nuclear Regulatory Commission's monthly Licensed Operating Reactors Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the errata page

  16. Licensed operating reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units is provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  17. Licensed operating reactors

    International Nuclear Information System (INIS)

    1989-06-01

    The US Nuclear Regulatory Commission's monthly LICENSED OPERATING REACTORS Status Summary Report provides data on the operation of nuclear units as timely and accurately as possible. This information is collected by the Office of Information Resources Management, from the Headquarters Staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities. Since all of the data concerning operation of the units are provided by the utility operators less than two weeks after the end of the month, necessary corrections to published information are shown on the ERRATA page. This report is divided into three sections: the first contains monthly highlights and statistics for commercial operating units, and errata from previously reported data; the second is a compilation of detailed information on each unit, provided by NRC Regional Offices, IE Headquarters and the Utilities; and the third section is an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non-power reactors in the United States

  18. The qualification of reactor operators

    International Nuclear Information System (INIS)

    Lima, J.M. de; Soares, H.V.

    1981-01-01

    The qualification and performance of nuclear power personnel have an important influence on the availability and safety operation of these plants. This paper describes the Brazilian rules and norms established by the CNEN-Brazilian Atomic Energy Comission, as well as policy of other countries concerning training requirements and experiences of nuclear power reactor operators. Some coments are made about the im pact of the march 1979 Three Mile Island accident on upgrading the reactor training requirements in U.S.A. and its international implication. (Author) [pt

  19. Licensing of nuclear reactor operators

    International Nuclear Information System (INIS)

    1979-09-01

    Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

  20. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    International Nuclear Information System (INIS)

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m 2 and a surface heat flux of 1 MW/m 2 . The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO 2 rods. The helium coolant pressure is 5 MPa, entering the module at 297 0 C and exiting at 550 0 C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter

  1. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1981-08-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  2. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    Anderson, T.V.

    1970-01-01

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  3. Method of reactor operation

    International Nuclear Information System (INIS)

    Nakajima, Takeshi

    1988-01-01

    Purpose: To minimize the power change due to the increase in xenone and power distribution after reaching the rated power in the case of using fresh fuels no requiring conditioning operation thereby starting the nuclear reactor in a short period of time and stably. Method: When control rods are entirely inserted only with a purpose for the compensation of the reactivity in a xenon-unsaturated state such as upon starting of the nuclear reactor, peaking is generated in the lower portion of the reactor core. Therefore, it is necessary to insert control rods for additionally suppressing the peaking in the lower portion of the reactor core to a relatively shallow level. In view of the above, a plurality of control rods are divided into a first control rod group finally inserted in the rated power state and a second control rod group other than the above. Then, the power is once elevated to the rated power level by means of such an intermediate control rod pattern that the ratio of the total extraction amount between the first control rod group and the second control rod group is made constant. Then, the control rods are extracted stepwise while setting the ratio of the total extraction amount constant in accordance with the change of the accumulating amount of xenone, to thereby obtain the purpose. (kamimura, M.)

  4. Licensed operating reactors

    International Nuclear Information System (INIS)

    1991-08-01

    The Nuclear Regulatory Commission's annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December because that report contains data for the month of December, the year to date (in this case calendar 1990) and cumulative data, usually from the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

  5. Licensed operating reactors

    International Nuclear Information System (INIS)

    Hartfield, R.A.

    1994-03-01

    The Nuclear Regulatory Commissions annual summary of licensed nuclear power reactor data is based primarily on the report of operating data submitted by licensees for each unit for the month of December, the year to date (in this case calendar year 1993) and cumulative data, usually for the date of commercial operation. The data is not independently verified, but various computer checks are made. The report is divided into two sections. The first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 capacity and availability factors are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensee and notes as to the use of weighted averages and starting dates other than commercial operation are provided

  6. Reactor operation method

    International Nuclear Information System (INIS)

    Suzuki, Toshio; Hida, Kazuki; Yoshioka, Ritsuo.

    1990-01-01

    The enrichment degree of fuels initially loaded in a reactor core was made extremely lower than that of fresh fuels to be loaded in the succeeding cycle, or the enrichment degree for all of the initially loaded fuels was made identical with that of the fresh fuels in the conventional reactor operation method. In this operation method, since the initially loaded fuels are sometimes taken out after the completion of the cycle at the low burnup degree as it is, it can not be said to reduce the fuel cycle cost. As a means for dissolving this problem, at least two different kinds of initially loaded fuels are prepared. The enrichment degree of the highly enriched fuels is made identical with that of the fresh fuels, and the enrichment degree and the number of low enriched fuels are not changed after the completion of the first cycle but they are operated till the end of the second cycle. Further, all of the fuels at the low enrichment degree are taken out after the completion of the second cycle and exchanged with the fresh fuels. As a result, high burnup ratio of the initially loaded fuels can be increased, to improve the fuel economy. (I.S.)

  7. Nuclear reactor operator licensing

    International Nuclear Information System (INIS)

    Bursey, R.J.

    1978-01-01

    The Atomic Energy Act of 1954, which was amended in 1974 by the Energy Reorganization Act, established the requirement that individuals who had the responsibility of operating the reactors in nuclear power plants must be licensed. Section 107 of the act states ''the Commission shall (1) prescribe uniform conditions for licensing individuals; (2) determine the qualifications of such individuals; and (3) issue licenses to such individuals in such form as the Commission may prescribe.'' The article discusses the types of licenses, the selection and training of individuals, and the administration of the Nuclear Regulatory Commission licensing examinations

  8. The Halden Reactor Project Workshop on Studies of Operator Performance During Night Shift's

    International Nuclear Information System (INIS)

    Morisseau, Dolores S.; Braarud, Per Oeyvind; Collier, Steve; Droeivoldsmo, Asgeir; Larsen, Marit; Lirvall, Peter

    1996-01-01

    A workshop on Studies of Operator Performance during Nights Shifts was organised in Halden, February 27-28, 1996. The purpose of the workshop was to discuss and make recommendations on specific needs for the study of operator cognitive performance at night and identify the relevant research issues for which Halden could provide resolution. The workshop began with presentations by several invited speakers with expertise in studies of shift work and was then divided into three working groups that discussed the following issues in parallel: (1) Lines of Research to Be Pursued; (2) Methods and Measures to Be Used in Research on Cognitive Performance at Nights; and (3) Products of the Research on Operator Performance at Night. Each group produced specific recommendations that were summarised by the group's facilitator in a joint session of the workshop. This report summarises the presentation of the invited speakers, and the discussions and recommendations of the individual working groups. (author)

  9. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-04-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis

  10. Performance evaluation of reactor operated zircaloy-2 pressure tubes of RAPS-1

    International Nuclear Information System (INIS)

    Chatterjee, S.; Ramadasan, E.; Balakrishnan, K.S.; Bahl, J.K.

    1992-01-01

    Detailed post irradiation examination was carried out on pressure tube sections from E-10, F-9 and F-10 locations of RAPS-1 after an in-reactor residence equivalent to 3.6 effective full power years. The F-10 pressure tube was studied in detail on sections obtained from one end to the other, whereas in the case of E-9 and F-9 pressure tubes only the end sections were examined. The studies carried out were visual examination, metallography, hydrogen i.e. H(D) analysis and mechanical testing at 300 C. Microstructural observations revealed uniform and random hydride/deuteride platelet distribution and absence of blisters or hydride segregation. The H(D) content in the F-10 pressure tube was found to vary in the range 6-12 ppm. The typical H(D) content in the three tubes was around 1 ppm. The H(D) pick-up evaluated from the observed oxide layer thickness was 8 ppm. Longitudinal tensile specimens fabricated from the F-10 pressure tube section and tested at 300 C exhibited increase in yield strength and tensile strength of 39% and 30% respectively. The residual uniform elongation was typically 1.8%. The observed changes in the tensile properties were found to be lower than those reported on unstressed specimens irradiated to similar neutron fluences. The observed hydrogen content and tensile properties obtained in F-10 pressure tube would not be detrimental under normal reactor operating conditions. (author). 10 refs., 4 figs., 2 tabs., 1 annexure

  11. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  12. Operational power reactor health physics

    International Nuclear Information System (INIS)

    Watson, B.A.

    1987-01-01

    Operational Health Physics can be comprised of a multitude of organizations, both corporate and at the plant sites. The following discussion centers around Baltimore Gas and Electric's (BG and E) Calvert Cliffs Nuclear Power Plant, located in Lusby, Maryland. Calvert Cliffs is a twin Combustion Engineering 825 MWe pressurized water reactor site with Unit I having a General electric turbine-generator and Unit II having a Westinghouse turbine-generator. Having just completed each Unit's ten-year Inservice Inspection and Refueling Outge, a total of 20 reactor years operating health physics experience have been accumulated at Calvert Cliffs. Because BG and E has only one nuclear site most health physics functions are performed at the plant site. This is also true for the other BG and E nuclear related organizations, such as Engineering and Quality Assurance. Utilities with multiple plant sites have corporate health physics entity usually providing oversight to the various plant programs

  13. CANDU reactor experience: fuel performance

    International Nuclear Information System (INIS)

    Truant, P.T.; Hastings, I.J.

    1985-07-01

    Ontario Hydro has more than 126 reactor-years experience in operating CANDU reactors. Fuel performance has been excellent with 47 000 channel fuelling operations successfully completed and 99.9 percent of the more than 380 000 bundles irradiated operating as designed. Fuel performance limits and fuel defects have had a negligible effect on station safety, reliability, the environment and cost. The actual incapability charged to fuel is less than 0.1 percent over the stations' lifetimes, and more recently has been zero

  14. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  15. Proven power reactor systems - novel features and developments in operation performance, safety and reliability

    International Nuclear Information System (INIS)

    Bugl, J.

    1975-01-01

    As the development of nuclear reactors for the generation of electric power started after the end of the Second World War, the prospective use of diverse materials as fuel, moderator and coolant resulted in a wide diversity of design possibilities. Of the 10 nuclear reactor types which were being considered most seriously in those days, only a few have achieved acceptance. This development is best illustrated by listing the nuclear power plants in service, under construction and on order at present, separately by reactor types (table). In the lead at present and for some years to come are the thermal reactors and especially the light water reactors (LWR). In the LWR group the lead is held by the pressurised water reactor (PWR) which accounts for 44% of the installed capacity of all the nuclear power plants in service at present. In the early 1980s this share will increase to 58%, whereas the share of the boiling water reactor (BWR) will increase to only 28% from 23% at present. (orig./TK) [de

  16. PUSPATI Triga Reactor - First year in operation

    International Nuclear Information System (INIS)

    Nahrul Khair Rashid.

    1983-01-01

    First year operation of RTP reactor was mostly devoted to making in house training, setting up and testing the facilities in preparation for more routine operations. Generally the operations are categorized into 4 main purposes; experiment of research, teaching and training, demonstration, and testing and maintenance. These four purposes are elaborated in detail. Additions and modifications were performed in order to improve the safety of reactor operation. (A.J.)

  17. Monitoring device for reactor operation

    International Nuclear Information System (INIS)

    Sakagami, Masaharu.

    1980-01-01

    Purpose: To increase the freedom for the power control due to control rod operation and flow rate control, as well as prevent fuel failures by the provision of a power distribution forecasting device for forecasting the changes in the reactor core power distribution and a device for calculating the fuel performance index and judging to display the calculated values. Constitution: The results for the calculation of the reactor core power distribution from a process computer that processes each of measuring signals of a nuclear power plant are used as inputs to a fuel power history calculator to constitute the power history up to the present time for each of the fuels. The date are inputted to a fuel performance index calculator to calculate the fuel performance index at present time for each of the fuels. Changes in the power distribution are forecast in a forecasting device for reactor power distribution relative to the changes in the control variables of a control variable memory unit and the date are inputted to a fuel power history calculator to forecast the power changes for each of the fuels. The amount of the power changes is inputted to a fuel performance index calculator and a fuel performance indicating and judging device judges and displays if they exceed a predetermined value. (Seki, T.)

  18. Method of operating a reactor

    International Nuclear Information System (INIS)

    Oosumi, Katsumi; Yamamoto, Michiyoshi.

    1980-01-01

    Purpose: To prevent stress corrosion cracking in the structural material of a reactor pressure vessel. Method: Prior to the starting of a reactor, the reactor pressure vessel is evacuated to carry out degassing of reactor water, and, at the same time, reactor water is heated. After reactor water is heated to a predetermined temperature, control rods are extracted to start nuclear heating. While the temperature of the reactor water is in a temperature range where elution of a metal which is a structural material of the reactor pressure vessel becomes vigorous and the sensitivity to the stress corrosion cracks increases, the reactor is operated at the maximum permissible temperature raising speed or maximum permissible cooling speed. (Aizawa, K.)

  19. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1983-01-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  20. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-05-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  1. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1983-03-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  2. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-07-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  3. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-11-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  4. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-10-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  5. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-08-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  6. Operating reactors licensing actions summary

    International Nuclear Information System (INIS)

    1982-09-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  7. Performance of operating and advanced light water reactor designs. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2001-10-01

    Nuclear power can provide security of energy supply, stable energy costs, and can contribute to greenhouse gas reduction. To fully realize these benefits, a continued and strong focus must be maintained on means for assuring the economic competitiveness of nuclear power relative to alternatives. Over the past several years, considerable improvements have been achieved in nuclear plant performance. Worldwide, the average energy availability factor has increased from 66 per cent in 1980 to 81 per cent in 1999, with some utilities achieving significantly higher values. This is being achieved through integrated programmes including personnel training and quality assurance, improvements in plant system and component design and plant operation, by various means to reduce outage duration for maintenance and refuelling and other scheduled shutdowns, and by reducing the number of forced outages. Application of technical means for achieving high performance of nuclear power plants is an important element for assuring their economic competitiveness. For the current plants, proper management includes development and application of better technologies for inspection, maintenance and repair. For future plants, the opportunity exists during the design phase to incorporate design features and technologies for achieving high performance. This IAEA Technical Committee meeting (TCM) provided a forum for information exchange on design features and technologies incorporated into LWR plants commissioned within the last 15-20 years, and into evolutionary LWR designs still under development, for achieving performance improvements with due regard to stringent safety requirements and objectives. It also addressed on-going technology development expected to achieve further improvements and/or significant cost reductions. The TCM was attended by 32 participants from 14 Member States: Argentina, Bulgaria, Czech Republic, Finland, France, Germany, Hungary, Japan, Republic of Korea, Mexico

  8. Nuclear analysis and performance of the Light Water Breeder Reactor (LWBR) core power operation at Shippingport

    International Nuclear Information System (INIS)

    Hecker, H.C.

    1984-04-01

    This report presents the nuclear analysis and discusses the performance of the LWBR core at Shippingport during power operation from initial startup through end-of-life at 28,730 EFPH. Core follow depletion calculations confirmed that the reactivity bias and power distributions were well within the uncertainty allowances used in the design and safety analysis of LWBR. The magnitude of the core follow reactivity bias has shown that the calculational models used can predict the behavior of U 233 -Th systems with closely spaced fuel rod lattices and movable fuel. In addition, the calculated final fissile loading is sufficiently greater than the initial fissile inventory that the measurements to be performed for proof-of-breeding evaluations are expected to confirm that breeding has occurred

  9. Reactor modification, preparation and operation

    International Nuclear Information System (INIS)

    Weill, J.; Furet, J.; Baillet, J.; Donvez, G.; Duchene, J.; Gras, R.; Mercier, R.; Chenouard, J.; Leconte, J.

    1962-01-01

    In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system

  10. Reactor modification, preparation and operation

    Energy Technology Data Exchange (ETDEWEB)

    Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

    1962-03-15

    In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

  11. Reactor modification, preparation and operation

    Energy Technology Data Exchange (ETDEWEB)

    Weill, J; Furet, J; Baillet, J; Donvez, G; Duchene, J; Gras, R; Mercier, R [Electronics Dept., Independent Section of Reactor Electronics, Saclay (France); Chenouard, J; Leconte, J [Dept. of Physical Chemistry, Stable Isotopes Section, Saclay (France)

    1962-03-01

    In the course of preparations for the dosimetry experiment at the R-B reactor the control and safety equipment of the reactor was found to be inadequate for operation at a constant power level of several watts. After completing the study of control and safety issues by CEA, safety and control were defined for the purpose of the Joint Dosimetry Experiment. Preparations for the Dosimetry Experiment included: installation of equipment for control and safety of the reactor; supplying 6570 Kg of heavy water by UK, reinforcement of the reactor wall on the outside of the building; constructing the protection of the control room; start-up, measuring of the critical heavy water level, and check of control and safety rods worth. After the final check of safety rod mechanisms, eight runs were performed at a power of 5 Watt, and then a 1 k Watt run was carried out and the power stabilized at this power for 30 min by automatic control system.

  12. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.; Katsumura, Y.; Yamada, K.; Shiga, S.; Moriya, K.; Yoshida, S.; Takahashi, H.

    2003-01-01

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  13. Simulator experiments: effects of experience of senior reactor operators and of presence of a shift technical advisor on performance in a boiling water reactor control room

    International Nuclear Information System (INIS)

    Beare, A.N.; Dorris, R.E.; Gray, L.H.

    1984-12-01

    This report describes the first experiment in a Nuclear Regulatory Commission-sponsored program of training simulator experiments and field data collection to evaluate the effects of selected performance shaping factors on the performance of nuclear power plant control room operators. The factors investigated were the experience level of the Senior Reactor Operator (SRO) and the presence of a Shift Technical Advisor (STA). Data were collected from 16 two-man crews of licensed operators (one SRO and one RO). The crews were split into high and low SRO-experience groups on the basis of the years of experience of the SROs as SROs. One half (4 of the 8 crews in each group) of the high- and low-SRO experience groups were assisted by an STA or an SRO acting as an STA. The crews responded to four simulated plant casualties which ranged in severity from an uncomplicated turbine trip to an anticipated transient without scram (ATWS). No significant differences in overall performance were found between groups led by high (25 to 114 months licensed as an SRO) and low (1 to 17 months as an SRO) experience SROs. However, crews led by low experience SROs tended to have shorter task performance times than crews led by high experience SROs. Although a tendency for the STA-assisted groups to score higher on four of the five measures was observed, the presence of the STA had no statistically significant effect on overall team performance. The correlation between individual performance, as measured by four of the task performance measures, and experience, measured by months as a licensed operator, was not statistically significant, nor was the correlation between task performance and recency of simulator training. 18 references, 5 figures, 13 tables

  14. Experience in operation of heavy water reactors

    International Nuclear Information System (INIS)

    Rotaru, Ion; Bilegan, Iosif; Ghitescu, Petre

    1999-01-01

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

  15. Influence of geometrical and operational parameters on the performance of porous catalytic membrane reactors

    NARCIS (Netherlands)

    Aran, H.C.; Klooster, H.J.G.; Jani, J.M.; Wessling, Matthias; Lefferts, Leonardus; Lammertink, Rob G.H.

    2012-01-01

    In this study, porous membrane reactors with various characteristic length (inner diameter), controllable catalyst support thickness, active catalyst surface area and tunable wetting properties are described for heterogeneously catalyzed gas¿liquid¿solid (G¿L¿S) reactions. We developed porous

  16. Reactor core operation management system

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Tomomi.

    1992-05-28

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.).

  17. Reactor core operation management system

    International Nuclear Information System (INIS)

    Sato, Tomomi.

    1992-01-01

    Among operations of periodical inspection for a nuclear power plant, sequence, time and safety rule, as well as necessary equipments and the number thereof required for each of the operation are determined previously for given operation plannings, relevant to the reactor core operations. Operation items relative to each of coordinates of the reactor core are retrieved and arranged based on specified conditions, to use the operation equipments effectively. Further, a combination of operations, relative to the reactor core coordinates with no physical interference and shortest in accordance with safety rules is judged, and the order and the step of the operation relevant to the entire reactor core operations are planned. After the start of the operation, the necessity for changing the operation sequence is judged depending on the judgement as to whether it is conducted according to the safety rule and the deviation between the plan and the result, based on the information for the progress of each of the operations. Alternatively, the operation sequence and the step to be changed are planned again in accordance with the requirement for the change of the operation planning. Then, the shortest operation time can be planned depending on the simultaneous operation impossible condition and the condition for the operation time zone determined by labor conditions. (N.H.)

  18. Evaluation of the RSG-GAS Alpha-Beta Aerosol Contaminant Monitor Performance Under Reactor Operation Condition

    International Nuclear Information System (INIS)

    Hartoyo, Unggul; Setiawanto, Anto; Sumarno, Yulius

    2000-01-01

    Analysis to evaluate the RSG-GAS alpha-beta aerosol contaminant monitor performance was done. The high potential radiation working area such as in RSG-GAS is important to monitored for personal safety. Further it is necessary to assure that the system monitor is reliable enough under normal conditions as well as emergency condition. The method uses in this analysis are monitoring and comparing with the standard source. The standard course indicator and panel in main control room indicate that the result is 1 x 110 exp 9 Ci/m exp 3. Based on data monitor observation, the RSG-GAS alpha-beta aerosol contaminant monitor system under reactor operation condition has a good enough performance

  19. Contribution of materials investigations and operating experience of reactor vessel internals to PWRs' safety, performance and reliability

    International Nuclear Information System (INIS)

    Lemaire, E.; Monteil, N.; Jardin, N.; Doll, M.

    2015-01-01

    The Reactor Pressure Vessel Internals (RVI) include all the components inside the pressure vessel, except the nuclear fuel, the rod cluster assemblies and the instrumentation. The RVI consist of bolted and welded structures that are divided into two sub-assemblies: the upper internals which are removed at every refueling outage and the lower internals which are systematically removed for inspection at every 10-year outage. The main functions of the RVI are to position the core, to support it, and to provide a coolant flow by channeling the fluid. Moreover, the lower internals contribute to a neutron protection of the reactor pressure vessel by absorbing most of the neutron flux from the core. Depending on their location and material composition, the RVI components can face different ageing phenomena, that are actual or potential (such as wear, fatigue, stress corrosion cracking, irradiation assisted stress corrosion cracking, hardening and loss of ductility due to neutron irradiation, irradiation creep and irradiation swelling). EDF has developed a strategy for managing ageing and demonstrating the capacity of the RVI to perform their design functions over 40 years of operation. This overall approach is periodically revisited to take into account the most recent knowledge obtained from the following main topics: Safety Analyses, Research-Development programs, In-Service Inspection (ISI) results, Maintenance programs and Metallurgical Examinations. Based on continuous improvements in those fields, the goal of this paper is to present the way that materials investigations and operating experience obtained on RVI are managed by EDF to improve RVI safety, performance and reliability. It is shown that a perspective of 60 years of operation of RVI components is supported by large Research-Development efforts combined with field experience. (authors)

  20. Reactor core performance calculating device

    International Nuclear Information System (INIS)

    Tominaga, Kenji; Bando, Masaru; Sano, Hiroki; Maruyama, Hiromi.

    1995-01-01

    The device of the present invention can calculate a power distribution efficiently at high speed by a plurality of calculation means while taking an amount of the reactor state into consideration. Namely, an input device takes data from a measuring device for the amount of the reactor core state such as a large number of neutron detectors disposed in the reactor core for monitoring the reactor state during operation. An input data distribution device comprises a state recognition section and a data distribution section. The state recognition section recognizes the kind and amount of the inputted data and information of the calculation means. The data distribution section analyzes the characteristic of the inputted data, divides them into a several groups, allocates them to each of the calculation means for the purpose of calculating the reactor core performance efficiently at high speed based on the information from the state recognition section. A plurality of the calculation means calculate power distribution of each of regions based on the allocated inputted data, to determine the power distribution of the entire reactor core. As a result, the reactor core can be evaluated at high accuracy and at high speed irrespective of the whole reactor core or partial region. (I.S.)

  1. Tendencies in operating power reactors

    International Nuclear Information System (INIS)

    Brinckmann, H.F.

    1987-01-01

    A survey is given about new tendencies in operating power reactors. In order to meet the high demands for control and monitoring of power reactors modern procedures are applicated such as the incore-neutron flux detection by means of electron emission detectors and multi-component activation probes, the noise diagnostics as well as high-efficient automation systems

  2. Experience in using a research reactor for the training of power reactor operators

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Arsenaut, L.J.

    1972-01-01

    A research reactor facility such as the one at the Omaha Veterans Administration Hospital would have much to offer in the way of training reactor operators. Although most of the candidates for the course had either received previous training in the Westinghouse Reactor Operator Training Program, had operated nuclear submarine reactors or had operated power reactors, they were not offered the opportunity to perform the extensive manipulations of a reactor that a small research facility will allow. In addition the AEC recommends 10 research reactor startups per student as a prerequisite for a cold operator?s license and these can easily be obtained during the training period

  3. Upgrade of reactor operation technology

    International Nuclear Information System (INIS)

    Itoh, Hideaki; Suzuki, Toshiaki; O-kawa, Toshikatsu

    2003-01-01

    To improve operational reliability and availability, the operation technology for a fast reactor was developed in the ''JOYO''. This report describes the upgrading of the simulator, plant operation management tools and fuel handling system for the MK-III core operation. The simulator was modified to the MK-III version to verify operation manuals, and to train operators in MK-III operation. The plant operation management tool was replaced on the operation experience to increase the reliability and efficiency of plant management works relating to plant operation and maintenance. To shorten the refueling period, the fuel handling system was upgraded to full automatic remote control. (author)

  4. RA reactor operation and maintenance

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-02-01

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air

  5. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

    1989-12-01

    This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

  6. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

    A vigorous commercial programme of isotope production and other radiation services has been followed by the SAFARI-1 research reactor over the past ten years - superimposed on the original purpose of the reactor to provide a basic tool for nuclear research, development and education to the country at an institutional level. A combination of the binding nature of the resulting contractual obligations and tighter regulatory control has demanded an equally vigorous programme of upgrading, replacement and renovation of many systems in order to improve the safety and reliability of the reactor. Not least among these changes is the more effective training and deployment of operations personnel that has been necessitated as the operational demands on the reactor evolved from five days per week to twenty four hours per day, seven days per week, with more than 300 days per year at full power. This paper briefly sketches the operational history of SAFARI-1 and then focuses on the training and structuring currently in place to meet the operational needs. There is a detailed step-by-step look at the operator?s career plan and pre-defined milestones. Shift work, especially the shift cycle, has a negative influence on the operator's career path development, especially due to his unavailability for training. Methods utilised to minimise this influence are presented. The increase of responsibilities regarding the operation of the reactor, ancillaries and experimental facilities as the operator progresses with his career are discussed. (author)

  7. DETAILS OF OPERATIONS PERFORMED BY THE REMOTE CONTROL ROBOT (CONCEPT TO THE HORIZONTAL FUEL CHANNEL DURING DECOMMISSIONING PHASE OF NUCLEAR REACTOR CALANDRIA STRUCTURE. PART II: INSIDE OPERATIONS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2017-05-01

    Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

  8. RBMK power unit operational performance investigation while false coming into action of the emergency reactor cooling system

    International Nuclear Information System (INIS)

    Emel'yanov, I.Ya.; Aleksakov, A.N.; Vasilevskij, V.P.; Labazov, V.N.; Nikolaev, E.V.; Podlazov, L.N.; Rogov, V.D.; Shevchenko, V.V.

    1984-01-01

    Regimes of RBMK reactor operation during false coming into action of the emergency reactor cooling system (ERCS), which might occur in the case of faults in the automation systems or erroneous actions of operator have been investigated. At that, thepe exists a probability of water supply from ERCS to one half of the reactor, which results in a sharp change of boiling regime, and due to void reactivity effect it causes the neutron field disturbance. Change in flow rate and enthalpy of coolant, as well as changes of neutron flux in the left and right halves of the reactor at ERCS response and during operation of the whole system of automatic control of power and system of local automatic control of power - local emergency protection - have been studied. The investigations have been carried out for different values of vapour effect of void reactivity effect and for time ranges from 0 to 40 s. The calculations are made using a model, describing spatial dynamics of reactor in two-dimensional approximation with 54 nodes. The model describes neutron-physics and thermohydraulic processes and it is realized using the BEhSM-6 computer. It is pointed out that one system of automatic control of power or local emergency protection (i.e. without the shut-off system), is insufficient for the compensation of disturbances appearing as a result of false ERCS coming into operation

  9. Operational and reliability experience with reactor instrumentation

    International Nuclear Information System (INIS)

    Dixon, F.; Gow, R.S.

    1978-01-01

    In the last 15 years the CEGB has experienced progressive plant development, integration and changes in operating regime through nine nuclear (gas-cooled reactor) power stations with corresponding instrumentation advances leading towards more refined centralized control. Operation and reliability experience with reactor instrumentation is reported in this paper with reference to the progressive changes related to the early magnox, late magnox and AGR periods. Data on instrumentation reliability in terms of reactor forced outages are presented and show that the instrumentation contributions to loss of generating plant availability are small. Reactor safety circuits, neutron flux and temperature measurements, gas analysis and vibration monitoring are discussed. In reviewing the reactor instrumentation the emphasis is on reporting recent experience, particularly on AGR equipment, but overall performance and changes to magnox equipment are included so that some appreciation can be obtained of instrumentation requirements with respect to plant lifetimes. (author)

  10. Method for operating nuclear reactor

    International Nuclear Information System (INIS)

    Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

    1978-01-01

    Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

  11. Example of a quality control operation performed on a nuclear reactor waste package

    International Nuclear Information System (INIS)

    Nouguier, H.; Nomine, J.S.; Vaunois, P.

    1983-01-01

    The need has emerged to secure the possibility of the spot sampling and inspection of a package, in the production line or on its arrival at the disposal center. This external inspection (or superinspection) has a threefold objective: to check the conformity of the package and of all its components with the mechanical, chemical and radiochemical specifications; to ensure, by thorough inspection, that the product is similar to the product which was characterized and qualified, and which secured ANDRA's approval of the process; and, through the detection of any variations and discrepancies, to advise the process promoter and the research laboratories, so as to improve the process, the product and the specifications. This paper describes an operation of this type conducted in 1982 by ANDRA, with the help of the CEA's laboratories and technical units, on low and medium-level waste packages produced by an industrial installation in an EDF nuclear power plant. The following are described: dismantling and sampling methods implemented, means employed, and the specific characteristics tested and the first results obtained

  12. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  13. Operational safety and reactor life improvements of Kyoto University Reactor

    International Nuclear Information System (INIS)

    Utsuro, M.; Fujita, Y.; Nishihara, H.

    1990-01-01

    Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

  14. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  15. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Preda, M.; Carabulea, A.

    2008-01-01

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

  16. Reactor Operations informal monthly report December 1994

    International Nuclear Information System (INIS)

    1994-12-01

    Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included

  17. Twenty years of health physics research reactor operation

    International Nuclear Information System (INIS)

    Sims, C.S.; Gilley, L.W.

    1983-01-01

    The Health Physics Research Reactor at the Oak Ridge National Laboratory has been in regular use for more than two decades. Safe operation of this fast reactor over this extended period indicates that (1) fundamental design, (2) operational procedures, (3) operator training and performance, (4) maintenance activites, and (5) management have all been eminently satisfactory. The reactor and its uses are described, the operational history and significant events are reviewed, and operational improvements and maintenance are discussed

  18. Design, construction and operating experience of demonstration LMFBRs. The application of core and fuel performance experience in British reactors to commercial fast reactor design

    International Nuclear Information System (INIS)

    Bagley, K.Q.

    1978-01-01

    The Prototype Fast Reactor (PFR) sub-assembly design is described with particular emphasis on the choice of factors that are important in determining satisfactory performance. Reasons for the adoption of specific clad and fuel design details are given in their historical context, and irradiation experience - mostly from the Dounreay Fast Reactor (DFR) - in support of the choices is described. The implications of factors that are now better understood than when the PFR fuel was designed, notably neutron-induced void swelling and irradiation creep, are then considered. It is shown that the 'free-standing' core design used in PFR, in which the sub-assembly is unsupported above the level of the lower axial breeder, relies on the availability of low-swelling, preferably irradiation-creep-resistant alloys as sub-assembly structural materials in order to achieve the prescribed burn-up target. The advantages of a 'restrained core', which makes use of irradiation creep to redress the effects of material swelling, are noted briefly, and the application of this concept to the Commercial Demonstration Fast Reactor (CDFR) core design is described. Probable future trends in pin and sub-assembly design are reviewed and the scope of associated irradiation testing programmes defined. Arrangements for monitoring and evaluating fuel performance, both in reactor and post-irradiation, are outlined and the provisions for endorsement of CDFR pin, sub-assembly and core design details in PFR are indicated. (author)

  19. Regulations and instructions for RA reactor operation

    International Nuclear Information System (INIS)

    1989-01-01

    This regulatory guide consists of following 4 chapters: Description of the RA reactor, organization scheme, regulations for performing experiments; Regulations for staff on duty; Instructions for operating the vacuum systems, heavy water and helium systems; and evacuation in case of accident [sr

  20. Reactor operator screening test experiences

    International Nuclear Information System (INIS)

    O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

    1976-01-01

    When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

  1. Health requirements for nuclear reactor operators

    International Nuclear Information System (INIS)

    1980-05-01

    The health prerequisites established for the qualification of nuclear reactor operators according to CNEN-NE-1.01 Guidelines Licensing of nuclear reactor operators, CNEN-12/79 Resolution, are described. (M.A.) [pt

  2. Operational safety evaluation for minor reactor accidents

    International Nuclear Information System (INIS)

    Wang, O.S.

    1981-01-01

    The purpose of this paper is to address a concern of applying conservatism in analysing minor reactor incidents. A so-called ''conservative'' safety analysis may exaggerate the system responses and result in a reactor scram tripped by the reactor protective system (RPS). In reality, a minor incident may lead the reactor to a new thermal hydraulic steady-state without scram, and the mitigation or termination of the incident may entirely depend on operator actions. An example on a small steamline break evaluation for a pressurized water reactor recently investigated by the staff at the Washington Public Power Supply System is presented to illustrate this point. A safety evaluation using mainly the safety-related systems to be consistent with the conservative assumptions used in the Safety Analysis Report was conducted. For comparison, a realistic analysis was also performed using both the safety- and control-related systems. The analyses were performed using the RETRAN plant simulation computer code. The ''conservative'' safety analysis predicts that the incident can be turned over by the RPS scram trips without operator intervention. However, the realistic analysis concludes that the reactor will reach a new steady-state at a different plant thermal hydraulic condition. As a result, the termination of the incident at this stage depends entirely on proper operator action. On the basis of this investigation it is concluded that, for minor incidents, ''conservative'' assumptions are not necessary, sometimes not justifiable. A realistic investigation from the operational safety point of view is more appropriate. It is essential to highlight the key transient indications for specific incident recognition in the operator training program

  3. Method of safely operating nuclear reactor

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro.

    1976-01-01

    Purpose: To provide a method of safely operating an nuclear reactor, comprising supporting a load applied to a reactor container partly with secondary container facilities thereby reducing the load borne by the reactor container when water is injected into the core to submerge the core in an emergency. Method: In a reactor emergency, water is injected into the reactor core thereby to submerge the core. Further, water is injected into a gap between the reactor container and the secondary container facilities. By the injection of water into the gap between the reactor container and the secondary container facilities a large apparent mass is applied to the reactor container, as a result of which the reactor container undergoes the same vibration as that of the secondary container facilities. Therefore, the load borne by the reactor container itself is reduced and stress at the bottom part of the reactor container is released. This permits the reactor to be operated more safely. (Moriyama, K.)

  4. Method of operating BWR type reactors

    International Nuclear Information System (INIS)

    Sekimizu, Koichi

    1980-01-01

    Purpose: To enable reactor control depending on any demanded loads by performing control by the insertion of control rods in addition to the control by the regulation of the flow rate of the reactor core water at high power operation of a BWR type reactor. Method: The power is reduced at high power operation by decreasing the flow rate of reactor core water from the starting time for the power reduction and the flow rate is maintained after the time at which it reaches the minimum allowable flow rate. Then, the control rod is started to insert from the above time point to reduce the power to an aimed level. Thus, the insufficiency in the reactivity due to the increase in the xenon concentration can be compensated by the withdrawal of the control rods and the excess reactivity due to the decrease in the xenon concentration can be compensated by the insertion of the control rods, whereby the reactor power can be controlled depending on any demanded loads without deviating from the upper or lower limit for the flow rate of the reactor core water. (Moriyama, K.)

  5. Fuel management for TRIGA reactor operators

    International Nuclear Information System (INIS)

    Totenbier, R.E.; Levine, S.H.

    1980-01-01

    One responsibility of the Supervisor of Reactor Operations is to follow the TRIGA core depletion and recommend core loading changes for refueling and special experiments. Calculations required to analyze such changes normally use digital computers and are extremely difficult to perform for one who is not familiar with computer language and nuclear reactor diffusion theory codes. The TRICOM/SCRAM program developed to perform such calculations for the Penn State TRIGA Breazeale Reactor (PSBR), has a very simple input format and is one which can be used by persons having no knowledge of computer codes. The person running the program need not understand computer language such as Fortran, but should be familiar with reactor core geometry and effects of loading changes. To further simplify the input requirements but still allow for all of the studies normally needed by the reactor operations supervisor, the options required for input have been isolated to two. Given a master deck of computer cards one needs to change only three cards; a title card, core energy history information card and one with core changes. With this input, the program can provide individual fuel element burn-up for a given period of operation and the k eff of the core. If a new loading is desired, a new master deck containing the changes is also automatically provided. The life of a new core loading can be estimated by feeding in projected core burn-up factors and observing the resulting loss in individual fuel elements. The code input and output formats have now been made sufficiently convenient and informative as to be incorporated into a standard activity for the Reactor Operations Supervisor. (author)

  6. Good practices in heavy water reactor operation

    International Nuclear Information System (INIS)

    2010-06-01

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

  7. Digital computer operation of a nuclear reactor

    International Nuclear Information System (INIS)

    Colley, R.W.

    1984-01-01

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state

  8. Reactor performance calculations for water reactors

    International Nuclear Information System (INIS)

    Hicks, D.

    1970-04-01

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

  9. Fast reactor parameter optimization taking into account changes in fuel charge type during reactor operation time

    International Nuclear Information System (INIS)

    Afrin, B.A.; Rechnov, A.V.; Usynin, G.B.

    1987-01-01

    The formulation and solution of optimization problem for parameters determining the layout of the central part of sodium cooled power reactor taking into account possible changes in fuel charge type during reactor operation time are performed. The losses under change of fuel composition type for two reactor modifications providing for minimum doubling time for oxide and carbide fuels respectively, are estimated

  10. Performance of a reactor containing denitrifying immobilized biomass in removing ethanol and aromatic hydrocarbons (BTEX) in a short operating period

    International Nuclear Information System (INIS)

    Gusmao, Valquiria Ribeiro; Chinalia, Fabio Alexandre; Sakamoto, Isabel Kimiko; Varesche, Maria Bernadete Amancio

    2007-01-01

    A horizontal-flow anaerobic immobilized biomass reactor (HAIB) containing denitrifying biomass was evaluated with respect to its ability to remove, separately and in a short operating period (30 days), organic matter, nitrate, and the hydrocarbons benzene (41.4 mg L -1 ), toluene (27.8 mg L -1 ), ethylbenzene (31.1 mg L -1 ), o-xylene (28.5 mg L -1 ), m-xylene (28.4 mg L -1 ) and p-xylene (32.1 mg L -1 ). The purified culture, which was grown in the presence of the specific hydrocarbon, was used as the source of cells to be immobilized in the polyurethane foam. After 30 days of operation, the foam was removed and a new immobilized biomass was grown in the presence of another hydrocarbon. The average hydrocarbon removal efficiency attained was 97%. The organic matter, especially ethanol, was removed with an average efficiency of 83% at a mean influent concentration of 1185.0 mg L -1 . A concomitant removal of 97% of nitrate was observed for a mean influent concentration of 423.4 mg L -1 . The independent removal of each hydrocarbon demonstrated that these contaminants can be biodegraded separately, without the need for a compound to be the primary substrate for the degradation of another. This study proposes the application of the system for treatment of areas contaminated with these compounds, with substitution and formation of a biofilm in a 30-day period

  11. Fractional power operation of tokamak reactors

    International Nuclear Information System (INIS)

    Mau, T.K.; Vold, E.L.; Conn, R.W.

    1986-01-01

    Methods to operate a tokamak fusion reactor at fractions of its rated power, identify the more effective control knobs and assess the impact of the requirements of fractional power operation on full power reactor design are explored. In particular, the role of burn control in maintaining the plasma at thermal equilibrium throughout these operations is studied. As a prerequisite to this task, the critical physics issues relevant to reactor performance predictions are examined and some insight into their impact on fractional power operation is offered. The basic tool of analysis consists of a zero-dimensional (0-D) time-dependent plasma power balance code which incorporates the most advanced data base and models in transport and burn plasma physics relevant to tokamaks. Because the plasma power balance is dominated by the transport loss and given the large uncertainty in the confinement model, the authors have studied the problem for a wide range of energy confinement scalings. The results of this analysis form the basis for studying the temporal behavior of the plasma under various thermal control mechanisms. Scenarios of thermally stable full and fractional power operations have been determined for a variety of transport models, with either passive or active feedback burn control. Important power control parameters, such as gas fueling rate, auxiliary power and other plasma quantities that affect transport losses, have also been identified. The results of these studies vary with the individual transport scaling used and, in particular, with respect to the effect of alpha heating power on confinement

  12. Performance and microbial community composition dynamics of aerobic granular sludge from sequencing batch bubble column reactors operated at 20 degrees C, 30 degrees C, and 35 degrees C.

    Science.gov (United States)

    Ebrahimi, Sirous; Gabus, Sébastien; Rohrbach-Brandt, Emmanuelle; Hosseini, Maryam; Rossi, Pierre; Maillard, Julien; Holliger, Christof

    2010-07-01

    Two bubble column sequencing batch reactors fed with an artificial wastewater were operated at 20 degrees C, 30 degrees C, and 35 degrees C. In a first stage, stable granules were obtained at 20 degrees C, whereas fluffy structures were observed at 30 degrees C. Molecular analysis revealed high abundance of the operational taxonomic unit 208 (OTU 208) affiliating with filamentous bacteria Leptothrix spp. at 30 degrees C, an OTU much less abundant at 20 degrees C. The granular sludge obtained at 20 degrees C was used for the second stage during which one reactor was maintained at 20 degrees C and the second operated at 30 degrees C and 35 degrees C after prior gradual increase of temperature. Aerobic granular sludge with similar physical properties developed in both reactors but it had different nutrient elimination performances and microbial communities. At 20 degrees C, acetate was consumed during anaerobic feeding, and biological phosphorous removal was observed when Rhodocyclaceae-affiliating OTU 214 was present. At 30 degrees C and 35 degrees C, acetate was mainly consumed during aeration and phosphorous removal was insignificant. OTU 214 was almost absent but the Gammaproteobacteria-affiliating OTU 239 was more abundant than at 20 degrees C. Aerobic granular sludge at all temperatures contained abundantly the OTUs 224 and 289 affiliating with Sphingomonadaceae indicating that this bacterial family played an important role in maintaining stable granular structures.

  13. Regulations and instructions for RB reactor operation; Propisi i uputstva za rad reaktora RB

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1977-07-01

    This document includes regulations for reactor RB operation, behaviour and presence of staff in the reactor building; regulations for performing experiments at the RB reactor, regulations and int ructions for the reactor operators and other staff on duty. A chapter is devoted to instruction for reactor operation with the operating documentation and special duties of the operators. Regulations and instruction concerned with accidents are described with classification of accidents and evacuation plan. Annexes to this document include: the present status of the reactor; program for training the reactor operators; forms which are obligatory to be signed for any operating activity, and the certificate of the RB reactor lattice.

  14. Reactor science and technology: operation and control of reactors

    International Nuclear Information System (INIS)

    Qiu Junlong

    1994-01-01

    This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

  15. Development of empirical models for performance evaluation of UASB reactors treating poultry manure wastewater under different operational conditions

    International Nuclear Information System (INIS)

    Yetilmezsoy, Kaan; Sakar, Suleyman

    2008-01-01

    A nonlinear modeling study was carried out to evaluate the performance of UASB reactors treating poultry manure wastewater under different organic and hydraulic loading conditions. Two identical pilot scale up-flow anaerobic sludge blanket (UASB) reactors (15.7 L) were run at mesophilic conditions (30-35 deg. C) in a temperature-controlled environment with three hydraulic retention times (θ) of 15.7, 12 and 8.0 days. Imposed volumetric organic loading rates (L V ) ranged from 0.65 to 4.257 kg COD/(m 3 day). The pH of the feed varied between 6.68 and 7.82. The hydraulic loading rates (L H ) were controlled between 0.105 and 0.21 m 3 /(m 2 day). The daily biogas production rates ranged between 4.2 and 29.4 L/day. High volumetric COD removal rates (R V ) ranging from 0.546 to 3.779 kg COD removed /(m 3 day) were achieved. On the basis of experimental results, two empirical models having a satisfactory correlation coefficient of about 0.9954 and 0.9416 were developed to predict daily biogas production (Q g ) and effluent COD concentration (S e ), respectively. Findings of this modeling study showed that optimal COD removals ranging from 86.3% to 90.6% were predicted with HRTs of 7.9, 9.5, 11.2, 12.6, 13.7 and 14.3 days, and L V of 1.27, 1.58, 1.78, 1.99, 2.20 and 2.45 kg COD/(m 3 day) for the corresponding influent substrate concentrations (S i ) of 10,000, 15,000, 20,000, 25,000, 30,000 and 35,000 mg/L, respectively

  16. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Wike, L.D.; Specht, W.L.; Mackey, H.E.; Paller, M.H.; Wilde, E.W.; Dicks, A.S.

    1989-12-01

    The Savannah River Site (SRS) is a large United States Department of Energy installation on the upper Atlantic Coastal Plain of South Carolina. The SRS contains diverse habitats, flora, and fauna. Habitats include upland terrestrial areas, varied wetlands including Carolina Bays, the Savannah River swamp system, and impoundment related and riparian wetlands, and the aquatic habitats of several stream systems, two large cooling reservoirs, and the Savannah River. These diverse habitats support a large variety of plants and animals including many commercially or recreational valuable species and several rare, threatened or endangered species. This volume describes the major habitats and their biota found on the SRS, and discuss the impacts of continued operation of the K, L, and P production reactors.

  17. Recent U.S. reactor operating experience

    International Nuclear Information System (INIS)

    Stello, V. Jr.

    1977-01-01

    A qualitative assessment of U.S. and foreign reactor operating experience is provided. Recent operating occurrences having potentially significant safety impacts on power operation are described. An evaluation of the seriousness of each of these issues and the plans for resolution is discussed. A quantitative report on U.S. reactor operational experience is included. The details of the NRC program for evaluating and applying operating reactor experience in the regulatory process is discussed. A review is made of the adequacy of operating reactor safety and environmental margins based on actual operating experience. The Regulatory response philosophy to operating reactor experiences is detailed. This discussion indicates the NRC emphasis on the importance of a balanced action plan to provide for the protection of public safety in the national interest

  18. Standards for safe operation of research reactors

    International Nuclear Information System (INIS)

    1996-01-01

    The safety of research reactors is based on many factors such as suitable choice of location, design and construction according to the international standards, it also depends on well trained and qualified operational staff. These standards determine the responsibilities of all who are concerned with the research reactors safe operation, and who are responsible of all related activities in all the administrative and technical stages in a way that insures the safe operation of the reactor

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  20. Light Water Reactor Sustainability Program Operator Performance Metrics for Control Room Modernization: A Practical Guide for Early Design Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Boring; Roger Lew; Thomas Ulrich; Jeffrey Joe

    2014-03-01

    As control rooms are modernized with new digital systems at nuclear power plants, it is necessary to evaluate the operator performance using these systems as part of a verification and validation process. There are no standard, predefined metrics available for assessing what is satisfactory operator interaction with new systems, especially during the early design stages of a new system. This report identifies the process and metrics for evaluating human system interfaces as part of control room modernization. The report includes background information on design and evaluation, a thorough discussion of human performance measures, and a practical example of how the process and metrics have been used as part of a turbine control system upgrade during the formative stages of design. The process and metrics are geared toward generalizability to other applications and serve as a template for utilities undertaking their own control room modernization activities.

  1. Fast reactor operating experience gained in Russia: Analysis of anomalies and abnormal operation cases

    International Nuclear Information System (INIS)

    Ashurko, Y.M.; Baklushin, R.P.; Zagorulko, Y.I.; Ivanenko, V.N.; Matveyev, V.P.; Vasilyev, B.A.

    2000-01-01

    Review of various anomalous events and abnormal operation experience gained in the process of Russian fast reactors operation is given in the paper. The main information refers to the BN-600 demonstration reactor operation. Statistical data on sodium leaks and steam generator failures are presented, and sources of these events and countermeasures taken to avoid their appearance on the operating reactors as well as related changes made in the BN-800 reactor design are considered. In the paper, some features of impurities behaviour are considered in various modes of the BN-600 reactor operation. Information is given on the impurities ingress into the circuits, on abnormal situation emerged in the process of the BN-600 reactor operation and its probable cause. Information is presented on the event related to the increased torque of the BN-600 reactor central rotating column and repair works performed. (author)

  2. Biogenic Hydrogen Conversion of De-Oiled Jatropha Waste via Anaerobic Sequencing Batch Reactor Operation: Process Performance, Microbial Insights, and CO2 Reduction Efficiency

    Directory of Open Access Journals (Sweden)

    Gopalakrishnan Kumar

    2014-01-01

    Full Text Available We report the semicontinuous, direct (anaerobic sequencing batch reactor operation hydrogen fermentation of de-oiled jatropha waste (DJW. The effect of hydraulic retention time (HRT was studied and results show that the stable and peak hydrogen production rate of 1.48 L/L*d and hydrogen yield of 8.7 mL H2/g volatile solid added were attained when the reactor was operated at HRT 2 days (d with a DJW concentration of 200 g/L, temperature 55°C, and pH 6.5. Reduced HRT enhanced the production performance until 1.75 d. Further reduction has lowered the process efficiency in terms of biogas production and hydrogen gas content. The effluent from hydrogen fermentor was utilized for methane fermentation in batch reactors using pig slurry and cow dung as seed sources. The results revealed that pig slurry was a feasible seed source for methane generation. Peak methane production rate of 0.43 L CH4/L*d and methane yield of 20.5 mL CH4/g COD were observed at substrate concentration of 10 g COD/L, temperature 30°C, and pH 7.0. PCR-DGGE analysis revealed that combination of celluloytic and fermentative bacteria were present in the hydrogen producing ASBR.

  3. dynamic performance of research reactors

    International Nuclear Information System (INIS)

    Abo elnor, A.G.M.

    2007-01-01

    this work studies the dynamic performance of material testing reactor (MTR), where the dynamic performance of any reactor reflects its safety behavior and it should enhance its intrinsic characteristics s ystem corrects itself internally without introducing external corrective action . the present work analyzes and studies the dynamic performance of mtr through the transfer function. the servo system parameters can be changed to fit the system demand. the servo system is an excellent approximation to some of the practical servo system currently use in reactor control system, and a quadratic form of this sort should closely approximate the behavior of almost any type of physical equipment which might be chosen to drive a control rod. proposed changes in servo system parameters could enhance the dynamic performance of the system , but the suitable parameters can be evaluated by using the automatic reactor power control system model

  4. Reactor shutdown: nuclear power plant performance

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    The article essentially looks at the performance of nine of Sweden's nuclear reactors. A table lists the percentage of time for the first three quarters of 1981 that the reactors were operating, and the number of hours out of service for planned or other reasons. In particular, one station - Ringhals 3 - was out of action because of a damaged tube in the associated steam generator. The same fault occurred with another reactor - Ringhals 4 - before this was brought into service. The reasons for the failure and its importance are briefly discussed. (G.P.)

  5. Regulations for RA reactor operation; Propisi nuklearnog reaktora 'RA'

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1980-09-15

    Regulations for RA reactor operation are written in accordance with the legal regulations defined by the Law about radiation protection and related legal acts, as well as technical standards according to the IAEA recommendations. The contents of this book include: fundamental data about the reactor; legal regulations for reactor operation; organizational scheme for reactor operation; general and detailed instructions for operation, behaviour in the reactor building, performing experiments; operating rules for operation under steady state and accidental conditions. [Serbo-Croat] Propisi o radu nuklearnog reaktora RA pisani su tako da svi zakonski propisi definisani 'Zakonom o zastiti od jonizujuceg zracenja' i pratecim propisima (devet pravilnika) kao i tehnicke norme prema preporukama MAAE budu postovani u punoj meri pri radu reaktora. Sadrzaj ove knjige obuhvata: osnovne podatke o reaktoru; zakonske propise; organizaciju rada reaktora RA; opste propise o rezimu rada, kretanju u zgradi reaktora, izvodjenju eksperimenata; pogonske propise za rad u normalnom rezimu i u slucaju udesa.

  6. Development of empirical models for performance evaluation of UASB reactors treating poultry manure wastewater under different operational conditions

    Energy Technology Data Exchange (ETDEWEB)

    Yetilmezsoy, Kaan [Department of Environmental Engineering, Yildiz Technical University, 34349 Yildiz, Besiktas, Istanbul (Turkey)], E-mail: yetilmez@yildiz.edu.tr; Sakar, Suleyman [Department of Environmental Engineering, Yildiz Technical University, 34349 Yildiz, Besiktas, Istanbul (Turkey)

    2008-05-01

    A nonlinear modeling study was carried out to evaluate the performance of UASB reactors treating poultry manure wastewater under different organic and hydraulic loading conditions. Two identical pilot scale up-flow anaerobic sludge blanket (UASB) reactors (15.7 L) were run at mesophilic conditions (30-35 deg. C) in a temperature-controlled environment with three hydraulic retention times ({theta}) of 15.7, 12 and 8.0 days. Imposed volumetric organic loading rates (L{sub V}) ranged from 0.65 to 4.257 kg COD/(m{sup 3} day). The pH of the feed varied between 6.68 and 7.82. The hydraulic loading rates (L{sub H}) were controlled between 0.105 and 0.21 m{sup 3}/(m{sup 2} day). The daily biogas production rates ranged between 4.2 and 29.4 L/day. High volumetric COD removal rates (R{sub V}) ranging from 0.546 to 3.779 kg COD{sub removed}/(m{sup 3} day) were achieved. On the basis of experimental results, two empirical models having a satisfactory correlation coefficient of about 0.9954 and 0.9416 were developed to predict daily biogas production (Q{sub g}) and effluent COD concentration (S{sub e}), respectively. Findings of this modeling study showed that optimal COD removals ranging from 86.3% to 90.6% were predicted with HRTs of 7.9, 9.5, 11.2, 12.6, 13.7 and 14.3 days, and L{sub V} of 1.27, 1.58, 1.78, 1.99, 2.20 and 2.45 kg COD/(m{sup 3} day) for the corresponding influent substrate concentrations (S{sub i}) of 10,000, 15,000, 20,000, 25,000, 30,000 and 35,000 mg/L, respectively.

  7. High performance light water reactor

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  8. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

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

  9. Computer monitoring of the RB reactor operation

    International Nuclear Information System (INIS)

    Milovanovic, S.; Pesic, M.; Milovanovic, T.

    1998-01-01

    Personal computer based acquisition system designed for monitoring of operation of the RB experimental reactor in the Institute of Nuclear Sciences 'Vinca' (former 'Boris Kidric') and experiences acquired during its use are shown in this paper. The monitoring covers generally all nuclear aspects of the reactor operation (start-up, nominal power operation, power changing, shut down and maintenance), but the emphasis is put on: real time (especially fast changing) reactivity measurement; supervising time dependence of the safety rods positions during shut down, and detection of position inaccuracy or failure operation of safety/control rods during the reactor operation or maintenance. (author)

  10. Operation and utilization of Indonesia Research Reactors

    International Nuclear Information System (INIS)

    Kuntoro, Iman; Sujalmo, Saiful; Tarigan, Alim

    2004-01-01

    For supporting the R and D in nuclear science and technology and its application, BATAN own and operate three research reactors namely, TRIGA-2000, KARTINI and RSG-GAS having thermal power of 2 MW, 100 kW and 30 MW respectively. The main features, operation and utilization progress of the reactors are described in this report. (author)

  11. LBB application in the US operating and advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  12. Safety of research reactors (Design and Operation)

    International Nuclear Information System (INIS)

    Dirar, H. M.

    2012-06-01

    The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)

  13. Operational performance of the development of a 15 kW parabolic trough mid-temperature solar receiver/reactor for hydrogen production

    International Nuclear Information System (INIS)

    Hong, Hui; Liu, Qibin; Jin, Hongguang

    2012-01-01

    Highlights: ► A 15 kW solar chemical receiver/reactor for hydrogen production was developed. ► The solar thermochemical efficiency of the receiver/reactor was in the range of 20–28%. ► Hydrogen production exceeding 80% was achieved. ► The research results extend the application of mid-temperature solar thermal energy. -- Abstract: In this paper, we report the operational performance and energy conversion efficiency of a developed 15 kW solar chemical receiver/reactor for hydrogen production. A concentrated solar heat of around 200–300 °C was utilized to provide process heat to drive methanol steam reforming. A modified 15 kW direct-irradiation solar reactor coupled with a linear receiver positioned along the focal line of a one-axis parabolic trough concentrator was used. The experiments were conducted from 200 to 300 °C under a mean solar flux of 300–800 W/m 2 and a reactant feeding rate of 6 kg/h. Reactants were continuously fed, and the attained conversion rate of methanol was more than 70% at 700 W/m 2 . The typical solar thermochemical efficiency of solar thermal energy converted into chemical energy was in the 20–28% range. The overall energy efficiency of input solar power conversion into chemical energy reached up to 17% and may be further increased by improving solar field efficiency. Hydrogen production exceeding 80% was achieved. In addition, preliminary economic evaluation was performed, and methods for further improvement were proposed. This paper proves that solar hydrogen production is feasible by combining solar thermal energy with alternative fuel at around 200–300 °C, which is much lower than the temperature of other solar thermochemical processes. This may offer an economic approach to solar fuel production and extend the application of mid-temperature solar thermal energy.

  14. Study on relationship of performance shaping factor in human error probability with prevalent stress of PUSPATI TRIGA reactor operators

    Science.gov (United States)

    Rahim, Ahmad Nabil Bin Ab; Mohamed, Faizal; Farid, Mohd Fairus Abdul; Fazli Zakaria, Mohd; Sangau Ligam, Alfred; Ramli, Nurhayati Binti

    2018-01-01

    Human factor can be affected by prevalence stress measured using Depression, Anxiety and Stress Scale (DASS). From the respondents feedback can be summarized that the main factor causes the highest prevalence stress is due to the working conditions that require operators to handle critical situation and make a prompt critical decisions. The relationship between the prevalence stress and performance shaping factors found that PSFFitness and PSFWork Process showed positive Pearson’s Correlation with the score of .763 and .826 while the level of significance, p = .028 and p = .012. These positive correlations with good significant values between prevalence stress and human performance shaping factor (PSF) related to fitness, work processes and procedures. The higher the stress level of the respondents, the higher the score of selected for the PSFs. This is due to the higher levels of stress lead to deteriorating physical health and cognitive also worsened. In addition, the lack of understanding in the work procedures can also be a factor that causes a growing stress. The higher these values will lead to the higher the probabilities of human error occur. Thus, monitoring the level of stress among operators RTP is important to ensure the safety of RTP.

  15. Reactor technology: power conversion systems and reactor operation and maintenance

    International Nuclear Information System (INIS)

    Powell, J.R.

    1977-01-01

    The use of advanced fuels permits the use of coolants (organic, high pressure helium) that result in power conversion systems with good thermal efficiency and relatively low cost. Water coolant would significantly reduce thermal efficiency, while lithium and salt coolants, which have been proposed for DT reactors, will have comparable power conversion efficiencies, but will probably be significantly more expensive. Helium cooled blankets with direct gas turbine power conversion cycles can also be used with DT reactors, but activation problems will be more severe, and the portion of blanket power in the metallic structure will probably not be available for the direct cycle, because of temperature limitations. A very important potential advantage of advanced fuel reactors over DT fusion reactors is the possibility of easier blanket maintenance and reduced down time for replacement. If unexpected leaks occur, in most cases the leaking circuit can be shut off and a redundant cooling curcuit will take over the thermal load. With the D-He 3 reactor, it appears practical to do this while the reactor is operating, as long as the leak is small enough not to shut down the reactor. Redundancy for Cat-D reactors has not been explored in detail, but appears feasible in principle. The idea of mobile units operating in the reactor chamber for service and maintenance of radioactive elements is explored

  16. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    Da Ruan; Benitez-Read, J.S.

    2005-01-01

    Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

  17. BEACON TSM application system to the operation of PWR reactors

    International Nuclear Information System (INIS)

    Lozano, J. A.; Mildrum, C.; Serrano, J. F.

    2012-01-01

    BEACON-TSM is an advanced core monitoring system for PWR reactor cores, and also offers the possibility to perform a wide range of predictive calculation in support of reactor operation. BEACON-TSM is presently installed and licensed in the 5 Spanish PWR reactors of standard Westinghouse design. the purpose of this paper is to describe the features of this software system and to show the advantages obtainable by a nuclear power plant from its use. To illustrate the capabilities and benefits of BEACON-TSM two real case reactor operating situations are presented. (Author)

  18. Method of operating a nuclear reactor

    International Nuclear Information System (INIS)

    Spurgin, A.J.; Schaefer, W.F.

    1978-01-01

    A method of controlling a nuclear power generting station in the event of a malfunction of particular operating components is described. Upon identification of a malfunction, preselected groups of control rods are fully inserted sequentially until a predetermined power level is approached. Additional control rods are then selectively inserted to quickly bring the reactor to a second given power level to be compatible with safe operation of the system with the malfunctioning component. At the time the thermal power output of the reactor is being reduced, the turbine is operated at a rate consistent with the output of the reactor. In the event of a malfunction, the power generating system is operated in a turbine following reactor mode, with the reactor power rapidly reduced, in a controlled manner, to a safe level compatible with the type of malfunction experienced

  19. Breakout Session B: Reactor Performance, R and D and Technological Issues. Breakout Session: Operation, R and D and technological issues

    International Nuclear Information System (INIS)

    Waeckel, Nicolas; Ott, Larry J.

    2013-01-01

    Three main outcomes resulted from this break-out session, as well as a few recommendations. Outcomes: 1 - The development of ATF is likely to be a long, stepwise/iterative approach, for which a broad consensus on the objectives is essential: - ATF should be as good as current fuels during normal operation ('first do no harm' principle), including good mechanical properties, and for DBA; - ATF should exhibit higher performances during severe accidents. 2 - For normal operation and DBA, the qualification path is well known and should follow the regulatory requirement: - the same as for an advanced fuel (properties characterisation, justification tests regarding SAFDLs of SRP4.2, etc.); - need to confirm though that the ATF candidate does not exhibit a specific behaviour or damage mechanisms that may require a specific justification (specific tests, specific SAFDL). 3 - For severe accidents, the recurrent questions are as follows: - What is the target? (in terms of expected performance); - What is the criterion for success? - Which tests should be run to justify the new fuel? Recommendations: 1 - In terms of organisation of the general development process: - identify the key factors or physical parameters that will really reduce the accident consequences; - calculate one BWR scenario (3-F) and one PWR (TMI-2) with the appropriate simulation tools; - run sensitivity analysis to scale the relative (beneficial) effect of various physical parameters (thermal conductivity, oxidation kinetics, H 2 production, exo-thermal reaction, FP retention, mechanical behaviour, etc.); - these calculations will be run with data progressively qualified on ad-hoc experience. 2 - International co-operation is recommended, especially during the feasibility phase: - no proprietary concern at this stage; - from utility point of view, it is better if all fuel vendors propose an ATF, to ensure the security of supply; - cross-irradiation of ATF candidates (ATR, HRP, JHR) can accelerate

  20. Reactor core performance estimating device

    International Nuclear Information System (INIS)

    Tanabe, Akira; Yamamoto, Toru; Shinpuku, Kimihiro; Chuzen, Takuji; Nishide, Fusayo.

    1995-01-01

    The present invention can autonomously simplify a neural net model thereby enabling to conveniently estimate various amounts which represents reactor core performances by a simple calculation in a short period of time. Namely, a reactor core performance estimation device comprises a nerve circuit net which divides the reactor core into a large number of spacial regions, and receives various physical amounts for each region as input signals for input nerve cells and outputs estimation values of each amount representing the reactor core performances as output signals of output nerve cells. In this case, the nerve circuit net (1) has a structure of extended multi-layered model having direct coupling from an upper stream layer to each of downstream layers, (2) has a forgetting constant q in a corrected equation for a joined load value ω using an inverse error propagation method, (3) learns various amounts representing reactor core performances determined using the physical models as teacher signals, (4) determines the joined load value ω decreased as '0' when it is to less than a predetermined value upon learning described above, and (5) eliminates elements of the nerve circuit net having all of the joined load value decreased to 0. As a result, the neural net model comprises an autonomously simplifying means. (I.S.)

  1. Operational experience of the Marcoule reactors

    International Nuclear Information System (INIS)

    Conte, F.

    1963-01-01

    The results obtaining from three years operation of the reactors G-2, G-3 have made it possible to accumulate a considerable amount of operational experience of these reactors. The main original points: - the pre-stressed concrete casing - the possibility of loading while under power - automatic temperature control have been perfectly justified by the results of operation. The author confirms the importance of these original solutions and draws conclusions concerning the study of future nuclear power stations. (author) [fr

  2. Manual for the operation of research reactors

    International Nuclear Information System (INIS)

    1965-01-01

    The great majority of the research reactors in newly established centres are light-water cooled and are often also light-water moderated. Consequently, the IAEA has decided to publish in its Technical Reports Series a manual dealing with the technical and practical problems associated with the safe and efficient operation of this type of reactor. Even though this manual is limited to light-water reactors in its direct application and presents the practices and experience at one specific reactor centre, it may also be useful for other reactor types because of the general relevance of the problems discussed and the long experience upon which it is based. It has, naturally, no regulatory character but it is hoped that it will be found helpful by staff occupied in all phases of the practical operation of research reactors, and also by those responsible for planning their experimental use. 23 refs, tabs

  3. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    2001-01-01

    There are three nuclear research reactors in the Czech Republic in operation now: light water reactor LVR-15, maximum reactor power 10 MW t , owner and operator Nuclear Research Institute Rez; light water zero power reactor LR-0, maximum reactor power 5 kW t , owner and operator Nuclear Research Institute Rez and training reactor VR-1 Sparrow, maximum reactor power 5 kW t , owner and operate Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. The training reactor VR-1 Vrabec 'Sparrow', operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly it is designed for training the students of Czech universities, preparing the experts for the Czech nuclear programme, as well as for certain research work, and for information programmes in the nuclear programme, as well as for certain research work, and for information programmes in sphere of using the nuclear energy (public relations). (author)

  4. Water Reactor Fuel Performance Meeting 2008

    International Nuclear Information System (INIS)

    2008-10-01

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

  5. Effect of biogas sparging on the performance of bio-hydrogen reactor over a long-term operation

    Science.gov (United States)

    Nualsri, Chatchawin; Kongjan, Prawit; Imai, Tsuyoshi

    2017-01-01

    This study aimed to enhance hydrogen production from sugarcane syrup by biogas sparging. Two-stage continuous stirred tank reactor (CSTR) and upflow anaerobic sludge blanket (UASB) reactor were used to produce hydrogen and methane, respectively. Biogas produced from the UASB was used to sparge into the CSTR. Results indicated that sparging with biogas increased the hydrogen production rate (HPR) by 35% (from 17.1 to 23.1 L/L.d) resulted from a reduction in the hydrogen partial pressure. A fluctuation of HPR was observed during a long term monitoring because CO2 in the sparging gas and carbon source in the feedstock were consumed by Enterobacter sp. to produce succinic acid without hydrogen production. Mixed gas released from the CSTR after the sparging can be considered as bio-hythane (H2+CH4). In addition, a continuous sparging biogas into CSTR release a partial pressure in the headspace of the methane reactor. In consequent, the methane production rate is increased. PMID:28207755

  6. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

  7. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

  8. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  9. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

  10. Preparation fo nuclear research reactors operators

    International Nuclear Information System (INIS)

    Roedel, G.

    1986-01-01

    The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

  11. Preparation of nuclear research reactors operators

    International Nuclear Information System (INIS)

    Roedel, G.

    1986-01-01

    The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN, are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

  12. Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

    International Nuclear Information System (INIS)

    Hauptman, H.M.; Petro, J.N.; Jacobi, O.

    1995-04-01

    This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

  13. Nuclear safety requirements for operation licensing of Egyptian research reactors

    International Nuclear Information System (INIS)

    Ahmed, E.E.M.; Rahman, F.A.

    2000-01-01

    From the view of responsibility for health and nuclear safety, this work creates a framework for the application of nuclear regulatory rules to ensure safe operation for the sake of obtaining or maintaining operation licensing for nuclear research reactors. It has been performed according to the recommendations of the IAEA for research reactor safety regulations which clearly states that the scope of the application should include all research reactors being designed, constructed, commissioned, operated, modified or decommissioned. From that concept, the present work establishes a model structure and a computer logic program for a regulatory licensing system (RLS code). It applies both the regulatory inspection and enforcement regulatory rules on the different licensing process stages. The present established RLS code is then applied to the Egyptian Research Reactors, namely; the first ET-RR-1, which was constructed and still operating since 1961, and the second MPR research reactor (ET-RR-2) which is now in the preliminary operation stage. The results showed that for the ET-RR-1 reactor, all operational activities, including maintenance, in-service inspection, renewal, modification and experiments should meet the appropriate regulatory compliance action program. Also, the results showed that for the new MPR research reactor (ET-RR-2), all commissioning and operational stages should also meet the regulatory inspection and enforcement action program of the operational licensing safety requirements. (author)

  14. Operation performance and granule characterization of upflow anaerobic sludge blanket (UASB) reactor treating wastewater with starch as the sole carbon source.

    Science.gov (United States)

    Lu, Xueqin; Zhen, Guangyin; Estrada, Adriana Ledezma; Chen, Mo; Ni, Jialing; Hojo, Toshimasa; Kubota, Kengo; Li, Yu-You

    2015-03-01

    Long-term performance of a lab-scale UASB reactor treating starch wastewater was investigated under different hydraulic retention times (HRT). Successful start-up could be achieved after 15days' operation. The optimal HRT was 6h with organic loading rate (OLR) 4g COD/Ld at COD concentration 1000mg/L, attaining 81.1-98.7% total COD removal with methane production rate of 0.33L CH4/g CODremoved. Specific methane activity tests demonstrated that methane formation via H2-CO2 and acetate were the principal degradation pathways. Vertical characterizations revealed that main reactions including starch hydrolysis, acidification and methanogenesis occurred at the lower part of reactor ("main reaction zone"); comparatively, at the up converting acetate into methane predominated ("substrate-shortage zone"). Further reducing HRT to 3h caused volatile fatty acids accumulation, sludge floating and performance deterioration. Sludge floating was ascribed to the excess polysaccharides in extracellular polymeric substances (EPS). More efforts are required to overcome sludge floating-related issues. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Management of operational events in research reactor

    International Nuclear Information System (INIS)

    Zhong Heping; Yang Shuchun; Peng Xueming

    2001-01-01

    The author describes the tracing management process post-operational event in a research reactor based on nuclear safety code, under the background of the research reactor in Nuclear Power Institute of China. It presorts the definite measures to the event tracing and it up its management factors

  16. Method of operating heavy water moderated reactors

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1980-01-01

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

  17. Fuel failure detection in operating reactors

    International Nuclear Information System (INIS)

    Seigel, B.; Hagen, H.H.

    1977-12-01

    Activity detectors in commercial BWRs and PWRs are examined to determine their capability to detect a small number of fuel rod failures during reactor operation. The off-gas system radiation monitor in a BWR and the letdown line radiation monitor in a PWR are calculated to have this capability, and events are cited that support this analysis. Other common detectors are found to be insensitive to small numbers of fuel failures. While adequate detectors exist for normal and transient operation, those detectors would not perform rapidly enough to be useful during accidents; in most accidents, however, primary system sensors (pressure, temperature, level) would provide adequate warning. Advanced methods of fuel failure detection are mentioned

  18. Impact of proposed research reactor standards on reactor operation

    Energy Technology Data Exchange (ETDEWEB)

    Ringle, J C; Johnson, A G; Anderson, T V [Oregon State University (United States)

    1974-07-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  19. Impact of proposed research reactor standards on reactor operation

    International Nuclear Information System (INIS)

    Ringle, J.C.; Johnson, A.G.; Anderson, T.V.

    1974-01-01

    A Standards Committee on Operation of Research Reactors, (ANS-15), sponsored by the American Nuclear Society, was organized in June 1971. Its purpose is to develop, prepare, and maintain standards for the design, construction, operation, maintenance, and decommissioning of nuclear reactors intended for research and training. Of the 15 original members, six were directly associated with operating TRIGA facilities. This committee developed a standard for the Development of Technical Specifications for Research Reactors (ANS-15.1), the revised draft of which was submitted to ANSI for review in May of 1973. The Committee then identified 10 other critical areas for standards development. Nine of these, along with ANS-15.1, are of direct interest to TRIGA owners and operators. The Committee was divided into subcommittees to work on these areas. These nine areas involve proposed standards for research reactors concerning: 1. Records and Reports (ANS-15.3) 2. Selection and Training of Personnel (ANS-15.4) 3. Effluent Monitoring (ANS-15.5) 4. Review of Experiments (ANS-15.6) 5. Siting (ANS-15.7) 6. Quality Assurance Program Guidance and Requirements (ANS-15.8) 7. Restrictions on Radioactive Effluents (ANS-15.9) 8. Decommissioning (ANS-15.10) 9. Radiological Control and Safety (ANS-15.11). The present status of each of these standards will be presented, along with their potential impact on TRIGA reactor operation. (author)

  20. RA Research reactor Annual report 1981 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Milosevic, M.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.

    1981-12-01

    The RA nuclear reactor stopped operation after March 1979 campaign due to appearance of aluminium oxyhydrates deposits on the surface of fuel element claddings. Relevant decisions of the Sanitary inspection body of the Ministry of health and the Director General of the 'Boris Kidric' Institute of nuclear sciences, Vinca, banned further reactor operation until reasons caused aluminium oxyhydrates deposition are investigated and removed to enable regular reactor operation. Until the end of 1979 and during 1980, after a series of analyses and findings that caused cease of reactor operation, all the preparatory actions needed for restart were performed. Due to the fact that there is no emergency cooling system and no appropriate filtering system at the reactor, and according to the new regulations about start up of nuclear facilities, the Sanitary inspection body made a decision about temporary licence for reactor start-up meaning performance of the 'zero experiment' limiting the operating power to 1% of the nominal power. Accordingly the reactor was restarted on January 21 1981. Criticality was reached with the core made of 80% enriched fuel elements only. After the experiment was finished by the end of March a permission was demanded for operation at higher power levels at full power. Taking into account the state of the reactor components the operating licence was issued limiting the power to 2 MW until reconstruction of the ventilation system and construction of the emergency cooling system are fulfilled. Program of testing operation started on September 15 1981 increasing gradually the operating power. Thus the reactor was operated at 2 MW power for 15 days during November and December. The total production achieved in 1981 was 1698 MWh. This enabled isotopes production at the reactor during last two months. Control and maintenance of the reactor components and systems was done regularly and efficiently within limits imposed by availability of spare parts. The

  1. 14th Biennial conference on reactor operating experience plant operations: The human element

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Separate abstracts were prepared for the papers presented in the following areas of interest: enhancing operator performance; structured approaches to maintenance standards and reliability-centered maintenance; human issues in plant operations and management; test, research, and training reactor utilization; methods and applications of root-cause analysis; emergency operating procedure enhancement programs; test, research, and training reactor upgrades; valve maintenance and diagnostics; recent operating experiences; and current maintenance issues

  2. Operation of the OSIRIS reactor from the viewpoint of analysis of operator functions

    International Nuclear Information System (INIS)

    Fichet-Clairfontaine, P.Y.; Saint-Jean, T.

    1985-09-01

    This paper presents the results of analyses carried out on site by the Human Factor Study Laboratory in an experimental nuclear plant operated by the Atomic Energy Commissariat - the OSIRIS pool reactor. The analyses of certain tasks are given: work in the reactor hall and an operation of circuit setting performed by the mechanics. This work has thrown light on certain operational guidelines implemented by the operators when carrying out their work [fr

  3. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  4. Operational reactor physics analysis codes (ORPAC)

    International Nuclear Information System (INIS)

    Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi

    2007-07-01

    For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)

  5. Report of the reactor Operators Service - Annex F

    International Nuclear Information System (INIS)

    Zivotic, Z.

    1992-01-01

    RA reactor operators service is organized in two groups: permanent staff (chief operator, chief shift operators and operators) and changeable group which is formed according to the particular operation needs for working in shifts. For continuous training of the existing operator staff the Service has prepared and published eleven booklets: Nuclear reactor; RA reactor primary coolant loop; System for purification of heavy water; reactor helium system; system for technical water; electric power system; control and operation; ventilation system in the reactor building; special sewage system; construction properties of the reactor core; reactor building and installations. During the reporting period there have been no accidents nor incidents that could affect the reactor personnel [sr

  6. Operational behaviour of a reactor normal operation and disturbances

    International Nuclear Information System (INIS)

    Geyer, K.H.

    1982-01-01

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

  7. The Effect of Influent Characteristics and Operational Conditions over the Performance and Microbial Community Structure of Partial Nitritation Reactors

    Directory of Open Access Journals (Sweden)

    Alejandro Rodriguez-Sanchez

    2014-06-01

    Full Text Available Nitrogen is a main contaminant of wastewater worldwide. Novel processes for nitrogen removal have been developed over the last several decades. One of these is the partial nitritation process. This process includes the oxidation of ammonium to nitrite without the generation of nitrate. The partial nitritation process has several advantages over traditional nitrification-denitrification processes for nitrogen removal from wastewaters. In addition, partial nitritation is required for anammox elimination of nitrogen from wastewater. Partial nitritation is affected by operational conditions and substances present in the influent, such as quinolone antibiotics. In this review, the impact that several operational conditions, such as temperature, pH, dissolved oxygen concentration, hydraulic retention time and solids retention time, have over the partial nitritation process is covered. The effect of quinolone antibiotics and other emerging contaminants are discussed. Finally, future perspectives for the partial nitritation process are commented upon.

  8. Independent verification: operational phase liquid metal breeder reactors

    International Nuclear Information System (INIS)

    Bourne, P.B.

    1981-01-01

    The Fast Flux Test Facility (FFTF) recently achieved 100-percent power and now is in the initial stages of operation as a test reactor. An independent verification program has been established to assist in maintaining stable plant conditions, and to assure the safe operation of the reactor. Independent verification begins with the development of administrative procedures to control all other procedures and changes to the plant configurations. The technical content of the controlling procedures is subject to independent verification. The actual accomplishment of test procedures and operational maneuvers is witnessed by personnel not responsible for operating the plant. Off-normal events are analyzed, problem reports from other operating reactors are evaluated, and these results are used to improve on-line performance. Audits are used to confirm compliance with established practices and to identify areas where individual performance can be improved

  9. Dynamics of nuclear reactor operational cycles

    International Nuclear Information System (INIS)

    Chapman, L.D.; Wayland, J.R.

    With this system dynamics computer model, one can explore the long term effects of a nuclear reactor program. Given an input demand for reactors, the consequences on each sector and the interactions among sectors can be simulated to provide a better understanding of the time development of a nuclear reactor program. The model permits the determination of various levels of activity as a function of time for plant enrichment, fuel fabrication, fuel reprocessing and storage of waste products. In addition, the rates of construction of reactors, spent fuel transit, disposal of waste, mining, shipping, recycling and enrichment can be investigated for optimal planning purposes. The model has been written in a very general manner so that it can be used to simulate any nuclear reactor program. It is an easy task to relate the amount of accidental or operational release of radioactive contaminants into our environment to the activity levels of each of the above sectors. (U.S.)

  10. Operational limits and conditions and operating procedures for research reactors. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This publication provides practical guidance on all important aspects of developing, formulating and presenting the operational limits and conditions as well as the operating procedures for research reactors. It covers the concept of operational limits and conditions, their content, and the responsibilities of the operating organization with respect to their establishment, modification, documentation and compliance. The guidance also covers the training of operating personnel on performing periodic testing, established by the operational limits and conditions, and operating procedures

  11. Reactor operation plan preparing device

    International Nuclear Information System (INIS)

    Sano, Hiroki; Maruyama, Hiromi; Kinoshita, Mitsuo; Fukuzaki, Koji; Banto, Masaru; Fukazawa, Yukihisa.

    1993-01-01

    The device comprises a means for retrieving a control rod pattern capable of satisfying a thermal limit upon aimed power/minimum flow rate and providing minimum xenon and a control rod pattern maximum xenon. It further comprises a means for selecting a control rod pattern corresponding to a xenon equilibrium condition, and selecting a control rod which provides a greater thermal margin to provide a control rod operation sequence for each of the patterns. Further, the device comprises an outline plan preparing means and a correction means therefor, a simplified sequence table reference means operated along with sequence change, an operation limit region input means, a control rod operation preferential region changing means, a thermal margin evaluation region and an input means. This can automatically prepare the operation plan, decrease the times for preparation of detailed plans by using the outline plan preparing function, thereby enabling to remarkably shorten the time for preparing of an operation plan. (N.H.)

  12. Reactor operation feed-back in France

    International Nuclear Information System (INIS)

    Feltin, C.; Fourest, B.; Libmann, J.

    1982-09-01

    The Nuclear Safety Department (DSN), technical support of French Safety Authorities, is, in particular, in charge of the analysis of reactor operation and of measures taken consequently to incidents. It proposed the criteria used to select significant incidents; it analyzes such incidents. DSN also analyzes the operating experience of each plant, several years after starting. It examines foreign incidents to assess in what extent lessons learned can be applied to french reactors. The examples presented show that to improve the safety of units operation, the experience feed-back leads to make arrangements, or modifications concerning not only circuits or materials but often procedures. Moreover they show the importance of procedures concerning the operations carried out during reactor shutdown

  13. A novel approach to the design and operation scheduling of heterogeneous catalytic reactors

    International Nuclear Information System (INIS)

    Ghodasara, Kamlesh; Smith, Robin; Hwang, Sungwon

    2014-01-01

    A number of studies have been conducted to reduce the overall level of catalyst deactivation in heterogeneous catalytic reactors, and improve the performance of reactors, such as yield, conversion or selectivity. The methodology generally includes optimization of the following: (1) operating conditions of the reaction system, such as feed temperature, normal operating temperature, pressure, and composition of feed streams; (2) reactor design parameters, such as dimension of the reactor, side stream distribution along the axis of the reactor beds, the mixing ratio of inert catalyst at each bed; and (3) catalyst design parameters, such as the pore size distribution across the pellet, active material distribution, size and shape of the catalyst, etc. Few studies have examined optimization of the overall catalyst reactor performance throughout the catalyst lifetime, considering catalyst deactivation. Furthermore, little attention has been given to the impact of various configurations of reactor networks and scheduling of the reactor operation (i.e., online and offline-regeneration) on the overall reactor performance throughout the catalyst lifetime. Therefore, we developed a range of feasible sequences of reactors and scheduling of reactors for operation and regeneration, and compared the overall reactor performance of multiple cases. Furthermore, a superstructure of reactor networks was developed and optimized to determine the optimum reactor network that shows the maximum overall reactor performance. The operating schedule of each reactor in the network was considered further. Lastly, the methodology was illustrated using a case study of the MTO (methanol to olefin) process

  14. Analysis of the radiometric survey during the Argonauta reactor operation

    International Nuclear Information System (INIS)

    Oliveira, Eara de S.L.; Cardozo, Katia K.M.; Silva, Joao Carlos P.; Santos, Joao Regis dos

    2013-01-01

    The Argonaut reactor at the Institute of Nuclear Engineering-IEN/CNEN, operates normally, the powers between 1.7 and 340 W on neutrongraphy procedures, production of radionuclides and experimental reactor physics lessons to postgraduate courses. The doses from neutrons and gamma radiation are measured when the reactor is critical, inside the reactor hall and surrounding regions. A study of the data obtained was performed to evaluate the daily need of this survey in the reactor hall. Taking into account the principle ALARA, which aims to optimize and minimize the dose received by the individual, we propose, in this work, through an analysis of the acquired data in occupational radiometric surveys, a reformulation of the area monitoring routine practiced by the team of radiological protection of the Institute of Nuclear Engineering - IEN/CNEN-RJ, whereas other monitoring routines regarding the radiological protection are also applied in the routine of the reactor. The operations under review occurred with the reactor operating 340 W power at intervals of 60, 120 and 180 minutes, in monitoring points in controlled areas, supervised and free. The results showed significant dose values in the output of the J-Channel 9 when the operation occurs with this open. With 180 minutes of operation, the measured values of dose rate were lower than the values at 60 min and 120 operations min. At the point in the supervised area, offsite to the reactor hall, situated in the direction of the J-Channel 9, the value reduces more than 14% in any operating time in relation to the dose rate measured at the point opposite the canal. There is a 50% reduction in the dose rates for operations with and J-9 closed. The results suggest a new frequency of radiometric survey whose mode of operation is maintained in similar conditions, since combined with other relevant practices of radiation protection

  15. Influence of substrates on nitrogen removal performance and microbiology of anaerobic ammonium oxidation by operating two UASB reactors fed with different substrate levels

    International Nuclear Information System (INIS)

    Tang Chongjian; Zheng Ping; Hu Baolan; Chen Jianwei; Wang Caihua

    2010-01-01

    Both ammonium and nitrite act as substrates as well as potential inhibitors of anoxic ammonium-oxidizing (Anammox) bacteria. To satisfy demand of substrates for Anammox bacteria and to prevent substrate inhibition simultaneously; two strategies, namely high or low substrate concentration, were carefully compared in the operation of two Anammox upflow anaerobic sludge blanket (UASB) reactors fed with different substrate concentrations. The reactor working at relatively low influent substrate concentration (NO 2 - N, 240 mg-N L -1 ) was shown to avoid the inhibition caused by nitrite and free ammonia. Using the strategy of low substrate concentration, a record super high volumetric nitrogen removal rate of 45.24 kg-N m -3 day -1 was noted after the operation of 230 days. To our knowledge, such a high value has not been reported previously. The evidence from transmission electron microscopy (TEM) showed that the morphology and ultrastructure of the Anammox cells in both the reactor enrichments was different.

  16. Physical-chemical and operational performance of an anaerobic baffled reactor (ABR treating swine wastewater - 10.4025/actascitechnol.v32i4.7203

    Directory of Open Access Journals (Sweden)

    Erlon Lopes Pereira

    2010-12-01

    Full Text Available Since hog raising concentrates a huge amount of swine manure in small areas, it is considered by the environmental government organizations to be one of the most potentially pollutant activities. Therefore the main objective of this research was to evaluate by operational criteria and removal efficiency, the performance of a Anaerobic Baffled Reactor (ABR, working as a biological pre-treatment of swine culture effluents. The physical-chemical analyses carried out were: total COD, BOD5, total solids (TS, fix (TFS and volatiles (TVS, temperature, pH, total Kjeldahl nitrogen, phosphorus, total acidity and alkalinity. The ABR unit worked with an average efficiency of 65.2 and 76.2%, respectively, concerning total COD and BOD5, with a hydraulic retention time (HRT about 15 hours. The results for volumetric organic loading rate (VOLR, organic loading rate (OLR and hydraulic loading rate (HLR were: 4.46 kg BOD m-3 day-1; 1.81 kg BOD5 kg TVS-1 day-1 and 1.57 m3 m-3 day-1, respectively. The average efficiency of the whole treatment system for total COD and BOD5 removal were 66.5 and 77.8%, showing an adequate performance in removing the organic matter from swine wastewater.

  17. Regulation for installation and operation of reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Concerning the description of an application for the approval of installation of a reactor, stipulated in Article 23 paragraph 2 of the Law for Regulation of Nuclear Source Materials, Nuclear Fuel Materials and Reactors (hereinafter referred to as the Law), the following items must be written. Namely, the heat output of the reactor in Article 23 paragraph 2 item 3 of the Law, the position, structure and facilities of the reactor facilities, described according to the stipulated classifications, the work plan, nuclear fuel materials employed, and the disposal of spent fuel. Concerning an application for the approval of a reactor installed aboard a foreign ship, stipulations are made separately. Description of an application for the approval of change of the heat output of a reactor and others should include the stipulated items. When it is wished to undergo inspection of the construction and performance of reactor facilities, an application for that end including the required items should be filed. Various safety measures preventing personnel from being exposed to radiation should be taken. When a foreign atomic-powered ship tries to enter a Japanese port, the stipulated necessary informations should be reported 60 days before such ship actually enters the Japanese port. A chief technician of reactors should take and pass the official examination. (Rikitake, Y.)

  18. Mode of operation of a nuclear reactor

    International Nuclear Information System (INIS)

    Morita, T.

    1976-01-01

    A method is proposed for the operation of a nuclear reactor guaranteeing an essentially symmetrical axial power distribution during normal operation by controlling the changes occuring in the reactor power partly by variation of the concentration of the neutron-absorbing element and partly by variation of the control rod positions. The representative parameters are recorded for the upper and lower half and adjusted to a predetermined reference value. In using this method, the axial power peals are reduced and power losses avoided. (RW) [de

  19. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  20. Performance indicators for power reactors

    International Nuclear Information System (INIS)

    Gillies, C.; White, M.

    1995-11-01

    A review of Canadian and worldwide performance indicator definitions and data was performed to identify a set of indicators that could be used for comparison of performance among nuclear power plants. The results of this review are to be used as input to an AECB team developing a consistent set of performance indicators for measuring Canadian power reactor safety performance. To support the identification of performance indicators, a set of criteria was developed to assess the effectiveness of each indicator for meaningful comparison of performance information. The project identified a recommended set of performance indicators that could be used by AECB staff to compare the performance of Canadian nuclear power plants among themselves, and with international performance. The basis for selection of the recommended set and exclusion of others is provided. This report provides definitions and calculation methods for each recommended performance indicator. In addition, a spreadsheet has been developed for comparison and trending for the recommended set of indicators. Example trend graphs are included to demonstrate the use of the spreadsheet. (author). 50 refs., 11 tabs., 3 figs

  1. Course of operators of the RA-3 reactor

    International Nuclear Information System (INIS)

    Caligiuri, G.A.

    1983-01-01

    Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

  2. The SM and MIR reactors operation experience

    International Nuclear Information System (INIS)

    Kuprienko, V.A.; Klinov, A.V.; Svyatkin, M.N.; Shamardin, V.K.

    1995-01-01

    The SM and MIR operation experience show that continuous work on the problem of ageing, in all its aspects, allows for prolongation of the research plant life cycle by several folds as compared to the initial project. The redesigned SM-3 reactor will operate for another 20 years. The similar result is expected from the MIR planned reconstruction which scope will be the topic of future presentations. (orig.)

  3. MAPLE research reactor beam-tube performance

    International Nuclear Information System (INIS)

    Lee, A.G.; Lidstone, R.F.; Gillespie, G.E.

    1989-05-01

    Atomic Energy of Canada Limited (AECL) has been developing the MAPLE (Multipurpose Applied Physics Lattice Experimental) reactor concept as a medium-flux neutron source to meet contemporary research reactor applications. This paper gives a brief description of the MAPLE reactor and presents some results of computer simulations used to analyze the neutronic performance. The computer simulations were performed to identify how the MAPLE reactor may be adapted to beam-tube applications such as neutron radiography

  4. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  5. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  6. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Tarigan, Alim

    1999-01-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  7. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  8. Operation and maintenance of the RB reactor, Annual report for 1978

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1978-01-01

    The annual report for 1978 includes the following: utilization of the RB reactor; producing the new safety report; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Four annexes to this report are concerned with: operation of the reactor at higher power levels; performance of the instrumentation, radiation doses during operation; gamma radiation doses after reactor shutdown; properties of the neutron converter (optimization of the rector-converter coupling; maintenance of the reactor equipment; purchase of new equipment

  9. RA reactor operation and maintenance in 1992, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Tanaskovic, M.

    1992-01-01

    During 1992 Ra reactor was not in operation. All the activities were fulfilled according to the previously adopted plan. Basic activities were concerned with revitalisation of the RA reactor and maintenance of reactor components. All the reactor personnel was busy with reconstruction and renewal of the existing reactor systems and building of the new systems, maintenance of the reactor devices. Part of the staff was trained for relevant tasks and maintenance of reactor systems [sr

  10. Method of controlling the reactor operation

    International Nuclear Information System (INIS)

    Ishiguro, Akira; Nakakura, Hiroyuki.

    1987-01-01

    Purpose: To moderate vibratory response due to delayed operation thereby obtain stable controlled response in the operation control for a PWR type reactor. Method: the reactor operation is controlled by the axial power distribution control by regulating the boron concentration in primary coolants with a boron density control system and controlling the average temperature for the primary coolants with the control rod control system. In this case, the control operation and the control response become instable due to transmission delay, etc. of aqueous boric acid injection to the primary coolant circuits to result in vibratory response. In the present invention, signals are prepared by adding the amount in proportion to the variation coefficient with time of xenone concentration obtained from the measured value for the reactor power added to the conventional axial power distribution parameter deviation and used as the input signals for the boron concentration control system. As a result, the instability due to the transmission delay of the aqueous boric acid injection is improved by the preceding control by the amount in proportion with the variation coefficient with time of the xenone concentration. An advantageous effect can be expected for the load following operation during day time according to the present invention. (Kamimura, M.)

  11. Operation of the High Flux Reactor. Annual report 1985

    International Nuclear Information System (INIS)

    1985-01-01

    This year was characterized by the end of a major rebuilding of the installation during which the reactor vessel and its peripheral components were replaced by new and redesigned equipment. Both operational safety and experimental use were largely improved by the replacement. The reactor went back to routine operation on February 14, 1985, and has been operating without problem since then. All performance parameters were met. Other upgrading actions started during the year concerned new heat exchangers and improvements to the reactor building complex. The experimental load of the High Flux Reactor reached a satisfactory level with an average of 57%. New developments aimed at future safety related irradiation tests and at novel applications of neutrons from the horizontal beam tubes. A unique remote encapsulation hot cell facility became available adding new possibilities for fast breeder fuel testing and for intermediate specimen examination. The HFR Programme hosted an international meeting on development and use of reduced enrichment fuel for research reactors. All aspects of core physics, manufacture technology, and licensing of novel, proliferation-free, research reactor fuel were debated

  12. Influence of substrates on nitrogen removal performance and microbiology of anaerobic ammonium oxidation by operating two UASB reactors fed with different substrate levels

    Energy Technology Data Exchange (ETDEWEB)

    Tang Chongjian [Department of Environmental Engineering, Zhejiang University, Hangzhou 310029 (China); Zheng Ping, E-mail: pzheng@zju.edu.cn [Department of Environmental Engineering, Zhejiang University, Hangzhou 310029 (China); Hu Baolan; Chen Jianwei; Wang Caihua [Department of Environmental Engineering, Zhejiang University, Hangzhou 310029 (China)

    2010-09-15

    Both ammonium and nitrite act as substrates as well as potential inhibitors of anoxic ammonium-oxidizing (Anammox) bacteria. To satisfy demand of substrates for Anammox bacteria and to prevent substrate inhibition simultaneously; two strategies, namely high or low substrate concentration, were carefully compared in the operation of two Anammox upflow anaerobic sludge blanket (UASB) reactors fed with different substrate concentrations. The reactor working at relatively low influent substrate concentration (NO{sub 2}{sup -}N, 240 mg-N L{sup -1}) was shown to avoid the inhibition caused by nitrite and free ammonia. Using the strategy of low substrate concentration, a record super high volumetric nitrogen removal rate of 45.24 kg-N m{sup -3} day{sup -1} was noted after the operation of 230 days. To our knowledge, such a high value has not been reported previously. The evidence from transmission electron microscopy (TEM) showed that the morphology and ultrastructure of the Anammox cells in both the reactor enrichments was different.

  13. 1984 Operation of the high flux reactor

    International Nuclear Information System (INIS)

    1985-01-01

    The programme resources in 1984 were largely devoted to the replacement of the old reactor vessel and its peripheral equipment. The original vessel had been in operation for more than 20 years and doubts had arisen about the condition of the aluminium tank after so long an exposure to neutrons. The operation, which had never been attempted before on a reactor of that size and complexity was planned and prepared over a number of years to take advantage of the occasion to provide a much improved vessel, incorporating the latest design features. The plant was shut down at the end of November 1983 and the 14 months operation began with a short cooling-off period for decay of short lived radioactivity followed by removal of the old tank and its dissection into pieces convenient for consolidation and storage as radioactive waste. After decontamination of the shielding pool, the new vessel and neutron beam tubes were installed and the reactor was recommissioned. Routine 45 MW operation was resumed on 14 February 1985 and has been uneventful since then

  14. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  15. Safe operation and maintenance of research reactor

    International Nuclear Information System (INIS)

    Munsorn, S.

    1999-01-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U 3 O 8 - Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  16. Operating reactors licensing actions summary. Vol.4, No. 4

    International Nuclear Information System (INIS)

    1984-06-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors

  17. Research reactor operations for neutron activation analysis

    International Nuclear Information System (INIS)

    Tv'ehlov, Yu.

    2002-01-01

    The IAEA Special Manual devoted to quality control during neutron activation analysis (NAA) on research and test reactors is discussed. Three parts of the publication involve presentation of common rules for performance of NAA, quantitative and qualitative analyses, statistic and systematic errors, safety regulations and radioactive waste management. Besides, the publication contains practical manual for the performance of NAA, and examples of different NAA regulating registration forms are presented [ru

  18. RA Research reactor, Annual report 1970 - Operation and maintenance

    International Nuclear Information System (INIS)

    Milosevic, D. et al.

    1970-12-01

    During 1970, the RA Reactor was operated at nominal power of 6.5 MW for 160 days, and 40 days at lower power levels. Total production mounted to 25968 MWh which is 3.87% higher than planned. The action plan was changed compared to the previous years because of sending the heavy water to France for re-concentration. Isotopic concentration of the heavy water was decreased to 99.05% and now after re-concentration it is 99.96%. Discrepancy from the action plan, in September was caused by the delay return of the heavy water for administrative and transportation difficulties. The restart of the reactor in September was postponed because the cladding of one fuel element was damaged immediately after the start-up, and the reactor had to be shutdown. In October and November reactor was in operation 28 and 25 days respectively which enabled to make up for the lost time. Reactor was used for irradiation and experiments according to the demand of 390 users, 340 from the Institute and 50 external users. This report contains detailed data about reactor power and experiments performed in 1969. It is concluded that the reactor operated successfully according to the plan. Shorter interruptions were caused only by difficulties with water supply pipes and sliding of the soil. Reactor was only twice scram shutdown because of the false signals caused by failures of the electronic control instrumentation. the period when reactor was not in operation was used for inspection of the reactor vessel internals. By using special TV cameras and telescopes, it was found that the there are no signs of corrosion on the reactor vessel, e.e. that the internals are in a very good state. Simultaneously, connection for the pipes of future emergency core cooling system were constructed. During 1970, the spent fuel was repacked from fuel channels into special aluminium casks. Four casks containing 660 fuel slugs was deposited int the storage pool No.4. There is now 18 casks with 2951 spent fuel slugs in

  19. Development of Advanced Monitoring System with Reactor Neutrino Detection Technique for Verification of Reactor Operations

    International Nuclear Information System (INIS)

    Furuta, H.; Tadokoro, H.; Imura, A.; Furuta, Y.; Suekane, F.

    2010-01-01

    Recently, technique of Gadolinium-loaded liquid scintillator (Gd-LS) for reactor neutrino oscillation experiments has attracted attention as a monitor of reactor operation and ''nuclear Gain (GA)'' for IAEA safeguards. When the thermal operation power is known, it is, in principle, possible to non-destructively measure the ratio of Pu/U in reactor fuel under operation from the reactor neutrino flux. An experimental program led by Lawrence Livermore National Laboratory and Sandia National Laboratories in USA has already demonstrated feasibility of the reactor monitoring by neutrinos at San Onofre Nuclear Power Station, and the Pu monitoring by neutrino detection is recognized as a candidate of novel technology to detect undeclared operation of reactor. However, further R and D studies of detector design and materials are still necessary to realize compact and mobile detector for practical use of neutrino detector. Considering the neutrino interaction cross-section and compact detector size, the detector must be set at a short distance (a few tens of meters) from reactor core to accumulate enough statistics for monitoring. In addition, although previous reactor neutrino experiments were performed at underground to reduce cosmic ray muon background, feasibility of the measurement at ground level is required for the monitor considering limited access to the reactor site. Therefore, the detector must be designed to be able to reduce external backgrounds extremely without huge shields at ground level, eg. cosmic ray muons and fast neutrons. We constructed a 0.76 ton Gd-LS detector, and carried out a reactor neutrino measurement at the experimental fast reactor JOYO in 2007. The neutrino detector was set up at 24.3m away from the reactor core at the ground level, and we understood the property of the main background; the cosmic-ray induced fast neutron, well. Based on the experience, we are constructing a new detector for the next experiment. The detector is a Gd

  20. A logic scheme for regulating safe operation of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Effat, A; Rahman, F A [Operational Saety Dept, National Center of Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    This investigation presents a logic scheme for regulating the safe operation of research reactor in accordance with the new revision of SS-35 and revised by the 10 CFR. It emphasizes the regulatory inspection and enforcement (RI end E) during the reactor operation phase. IT is developed to provide information, guidance and recommendations to be taken when constructing the RI and E program that could be applied to the operational phase of the egyptian Research Reactors. In the operational phase, the regulatory inspection (RI) means an examination, observation, measurement, or test undertaken or on behalf of the nuclear regulatory body (NRB) during operation to verify that the nuclear materials, components, systems and structures as well operational activities, processes, procedures and personnel competence and performance are in accordance with the requirements established or the provisions approved by NRB or specified in the operational license or contained in regulations. Regulatory inspection includes both routine and non-routine ones. Any of them may be announced or unannounced. The problems identified by the RI must be resolved by the proper RE actions. The RE actions include investigative and corrective RE actions. These RI and E procedures for regulating safe operation of research reactors are presented as flow charts and then developed as a computer logic scheme. The software program is very efficient, very friendly, very simple and is interactive in nature such that the program asks the user certain questions about essential steps that guide the (RI and E) for research reactors, and user responds. The program proceeds based on this response until all the necessary steps for (RI and E) are accomplished. 5 figs.

  1. RA reactor operation and maintenance, Annual report 1974

    International Nuclear Information System (INIS)

    Milosevic, D. et al

    1974-12-01

    During 1974, RA reactor was operated at nominal power for 194 days and 13 days at lower power levels. The total production was 30711 MWh which is 2.4% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 437 users. This report contains detailed data about reactor power and experiments performed in 1974. Total number of afety shutdowns was 11, of which 8 were caused by power cuts, and 3 due to human error. Maximum individual personnel exposure dose was 50% of the maximum permissible dose. There were no accidents during this year. Decontamination of surfaces was less than during previous years. About 805 m 2 of surfaces and 178 objects were decontaminated. It was concluded that the successful operation in 1974 has a special significance taking into account the financial problems [sr

  2. Review of Operation and Maintenance Support Systems for Research Reactors

    International Nuclear Information System (INIS)

    Jin, Kyungho; Heo, Gyunyoung; Park, Jaekwan

    2014-01-01

    Operation support systems do not directly control the plant but it can aid decision making itself by obtaining and analyzing large amounts of data. Recently, the demand of research reactor is growing and the need for operation support systems is increasing, but it has not been applied for research reactors. This study analyzes operation and maintenance support systems of NPPs and suggests appropriate systems for research reactors based on analysis. In this paper, operation support systems for research reactors are suggested by comparing with those of power reactors. Currently, research reactors do not cover special systems in order to improve safety and operability in comparison with power reactors. Therefore we expect to improve worth to use by introducing appropriate systems for research reactors. In further research, we will develop an appropriate system such as applications or tools that can be applied to the research reactor

  3. Graphics and control for in-reactor operations

    International Nuclear Information System (INIS)

    Smith, A.L.

    1996-01-01

    Remotely operated inspection, repair and maintenance tasks for Magnox Electric's eight twin MAGNOX-type reactor stations must conform to the Nuclear Installations Inspectorate's regulations. The performance of manipulator systems used to undertake such operations has improved significantly with the recent introduction of computer graphics. These are used for simulation and demonstration purposes as well as a real-time aid to the manipulator operator. Outage times can be reduced by the appropriate use of such computer technology to reduce manipulator operation times. (UK)

  4. Experimental evaluation of an expert system for nuclear reactor operators

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1984-10-01

    The United States Nuclear Regulatory Commission (USNRC) is supporting a program for the experimental evaluation of an expert system for nuclear reactor operators. A prototype expert system, called the Response Tree System, has been developed and implemented at INEL. The Response Tree System is designed to assess the status of a reactor system following an accident and recommend corrective actions to reactor operators. The system is implemented using color graphic displays and is driven by a computer simulation of the reactor system. Control of the system is accomplished using a transparent touch panel. Controlled experiments are being conducted to measure performance differences between operators using the Response Tree System and those not using it to respond to simulated accident situations. This paper summarizes the methodology and results of the evaluation of the Response Tree System, including the quantitative results obtained in the experiments thus far. Design features of the Response Tree System are discussed, and general conclusions regarding the applicability of expert systems in reactor control rooms are presented

  5. Reactor operations for nuclear pumping of lasers

    Energy Technology Data Exchange (ETDEWEB)

    Beck, G; Cooper, G [University of Illinois (United States)

    1974-07-01

    Experiments involving the measurement of gas parameters that are related to lasing, and lasing of various gas mixtures have comprised a major part of the utilization of the University of Illinois Advanced TRIGA Reactor since the upgrading of the facility was completed in 1969. A thru beam port, which was added during upgrading, has been the facility used for these measurements. The laser cell is placed in the port adjacent to the core. Alignment is then accomplished by using both ends of the port or by a mirror placed at the back side of the apparatus. The reactor has been operated in all modes (pulsing, square wave, and steady state) for the experiments although pulsing is the primary mode that is used. Laser enhancement has been obtained in several cases, but efforts toward direct pumping from the radiation alone have not as yet succeeded. Improved laser operation from direct pumping has been suggested with an emphasis on high-powered systems where the basic input energy is to be derived from a nuclear reactor.

  6. Fast reactor operation in the United States

    International Nuclear Information System (INIS)

    Smith, R.R.; Cissel, D.W.

    1978-01-01

    Of the many American facilities dedicated to fast reactor technology, six qualify as liquid-metal-cooled fast reactors. All of these satisfy the following criteria: an unmoderated neutron spectrum, highly enriched fuel material, substantial heat production, and the use of a liquid metal coolant. These include the following: EBR-I Clementine, LAMPRE, EBR-II, EFFBR, and SEFOR. Collectively, these facilities encompassed all of the more important features of liquid-metal-cooled fast reactor technology. Coolant types ranged from mercury in Clementine, to NaK in EBR-I, and sodium in the others. Fuels included enriched-uranium metallic alloys in EBR-I, EBR-II, and EFFBR; metallic plutonium in Clementine; molten plutonium alloy in LAMPRE; and a mixed UO 2 -PuO 2 ceramic in SEFOR. Heat removal techniques ranged from air-blast cooling in LAMPRE and SEFOR; steam-electrical generation in EBR-I, EBR-II, and EFFBR; to a mercury-to-water heat dump in Clementine. Operational experience with such diverse systems has contributed heavily to the U.S. Each of the six systems is described from the viewpoints of purpose, history, design, and operation. Attempts are made to limit descriptive material to the most important features and to refer the reader to a few select references if additional information is needed

  7. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  8. Control of water chemistry in operating reactors

    International Nuclear Information System (INIS)

    Riess, R.

    1997-01-01

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

  9. Control of water chemistry in operating reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  10. Linguistic Formalism for Semi-Autonomous Reactor Operation

    International Nuclear Information System (INIS)

    Joo, Sungmoon; Seo, Sang Mun; Suh, Yong-Suk; Park, Cheol

    2017-01-01

    The ultimate goal of our work is to develop a novel, integrated system for semi-autonomous reactor operation by introducing an interfacing language shared by human reactor operators and artificially intelligent service agents (e.g., robots). We envision that human operators and artificially intelligent service agents operate the reactor cooperatively in the future. For example, an artificially intelligent service agent carries out a human reactor operator's command or reports the result of a task commanded by the human reactor operator. This work presents preliminary work towards a unified linguistic formalism for cooperative, semiautonomous reactor operation. Application of the proposed formalism to reactor operator communication domain shows that the formalism effectively captures the syntax and semantics of the domain-specific language defined by the communication protocol.

  11. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    Hien, P.Z.

    1988-01-01

    The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

  12. Safety evaluation of the Dalat research reactor operation

    International Nuclear Information System (INIS)

    Long, V.H.; Lam, P.V.; An, T.K.

    1989-01-01

    After an introduction presenting the essential characteristics of the Dalat Nuclear Research Reactor, the document presents i) The safety assurance condition of the reactor, ii) Its safety behaviour after 5 years of operation, iii) Safety research being realized on the reactor. Following is questionnaire of safety evaluation and a list of attachments, which concern the reactor

  13. Application of SAFE to an operating reactor

    International Nuclear Information System (INIS)

    Chapman, L.D.

    1979-01-01

    A method for the evaluation of physical protection systems at nuclear facilities has been developed. The evaluation process consists of five major phases: (1) Facility Characterization, (2) Facility Representation, (3) Component Performance, (4) Adversary Path Analysis, and (5) Effectiveness Evaluation. Each of these phases will be described in some detail and illustrated by examples. The process for evaluation of physical protection system effectiveness against an outside threat will be presented for a reactor facility

  14. Operating reactors licensing actions summary. Vol. 3, No. 3

    International Nuclear Information System (INIS)

    1983-04-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  15. Operating reactors licensing actions summary. Volume 5, No. 2

    International Nuclear Information System (INIS)

    1985-04-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the Operating Reactors Licensing Actions Program

  16. Research and development into power reactor fuel performance

    International Nuclear Information System (INIS)

    Notley, M.J.F.

    1983-07-01

    The nuclear fuel in a power reactor must perform reliably during normal operation, and the consequences of abnormal events must be researched and assessed. The present highly reliable operation of the natural UO 2 in the CANDU power reactors has reduced the need for further work in this area; however a core of expertise must be retained for purposes such as training of new staff, retaining the capability of reacting to unforeseen circumstances, and participating in the commercial development of new ideas. The assessment of fuel performance during accidents requires research into many aspects of materials, fuel and fission product behaviour, and the consolidation of that knowledge into computer codes used to evaluate the consequences of any particular accident. This work is growing in scope, much is known from out-reactor work at temperatures up to about 1500 degreesC, but the need for in-reactor verification and investigation of higher-temperature accidents has necessitated the construction of a major new in-reactor test loop and the initiation of the associated out-reactor support programs. Since many of the programs on normal and accident-related performance are generic in nature, they will be applicable to advanced fuel cycles. Work will therefore be gradually transferred from the present, committed power reactor system to support the next generation of thorium-based reactor cycles

  17. PCI. Mechanism, measures, rules for reactor operation

    International Nuclear Information System (INIS)

    Bender, D.; Bender, G.; Dewes, P.; Wensauer, A.

    2009-01-01

    Though modern and advanced fuel assembly designs for BWR as AREVA's ATRIUM trademark 10 fuel assemblies show a high reliability, fuel failures are still encountered. The ZTF (Zero Tolerance for Failures) initiative was launched by AREVA to further upgrade BWR fuel assembly reliability towards failure free fuel. The introduction of the zirconium liner cladding has greatly reduced pellet cladding interaction (PCI) failures. However the phenomenon of PCI in reactor operation was not completely eliminated. There were failures in AREVA's BWR fuel with liner cladding attributed to PCI, but when subject to hot cell examinations, all these have been attributed to ''non-classical'' PCI, where local cladding stress is amplified due to missing pellet surface (MPS). However, in order to support the efforts to eliminate all possible root causes for fuel failures, AREVA NP addresses PCI with a ''multi-track'' strategy. First, and most important, the development of chromia doped pellets and more favorable pellet geometries aim at fuel that by design is protected as much as possible against PCI. A high pellet quality is further assured by the final assessment of the pellet appearance by an automatic visual inspection system in the pellet manufacturing lines. In order to support these hardware improvements, AREVA NP provides software aimed to minimize PCI risks. Two main approaches are described in this paper: - The ''classical'' PCI operation and maneuvering guidelines restrict power maneuvering and limit the ramping rates with the goal to control the stress level of the cladding. AREVA's guidelines are based on ramp tests as well as on extensive power plant experience, and are updated to reflect the performance of liner cladding. - Additional effort is made to further understand the PCI phenomenon, e.g. by the use of detailed fuel rod thermal-mechanical codes that provide far better insight of the state variables that control the process of rupture due to stress corrosion

  18. Operating experiences at the Finnish TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    1988-01-01

    The Finnish TRIGA reactor has been in operation since March 1962. There are still 57 original Al-clad fuel elements in the core. So far we have had only two fuel cladding failures in 1981 and 1988. The first one was an Al-clad element and the second one a SS-clad. The low rate of fuel cladding failures has made it possible to use continuously also the Al-clad fuel elements. Although some conventional irradiations of certain type have been repeated successfully tens of times, new and unexpected incidents can still take place. As an example an event of a leaking irradiation capsule is described

  19. Lessons from feedback of safety operating experience for reactor physics

    International Nuclear Information System (INIS)

    Suchomel, J.; Rapavy, S.

    1999-01-01

    Analyses of events in WWER operations as a part of safety experience feedback provide a valuable source of lessons for reactor physics. Examples of events from Bohunice operation will be shown such as events with inadequate approach to criticality, positive reactivity insertions, expulsion of a control rod from shut-down reactor, problems with reactor protection system and control rods. (Authors)

  20. Operating reactors licensing actions summary. Volume 5, No. 7

    International Nuclear Information System (INIS)

    1985-09-01

    The Operating Reactors Licensing Actions Summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  1. Operational methods of the fluidized bed nuclear reactor

    International Nuclear Information System (INIS)

    Borges, V.; Sefidvash, F.

    1993-01-01

    The operational curve of reactivity as a function of porosity of the Fluidized Bed Nuclear Reactor is presented. The strategies for start-up, shut-down and maintaining the reactor critical during operation are described. The inherent safety of the reactor from neutronic point of view under steady state condition is demonstrated. (author)

  2. Issues and approaches in control for autonomous reactor operation

    International Nuclear Information System (INIS)

    Vilim, R. B.; Khalil, H. S.; Wei, T. Y. C.

    2000-01-01

    A capability for autonomous and passively safe operation is one of the goals of the NERI funded development of Generation IV nuclear plants. An approach is described for evaluating the effect of increasing autonomy on safety margins and load behavior and for examining issues that arise with increasing autonomy and their potential impact on performance. The method provides a formal approach to the process of exploiting the innate self-regulating property of a reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error. Some preliminary results are given

  3. Space reactor fuels performance and development issues

    International Nuclear Information System (INIS)

    Wewerka, E.M.

    1984-01-01

    Three compact reactor concepts are now under consideration by the US Space Nuclear Power Program (the SP-100 Program) as candidates for the first 100-kWe-class space reactor. Each of these reactor designs puts unique constraints and requirements on the fuels system, and raises issues of fuel systems feasibility and performance. This paper presents a brief overview of the fuel requirements for the proposed space reactor designs, a delineation of the technical feasibility issues that each raises, and a description of the fuel systems development and testing program that has been established to address key technical issues

  4. Contour analysis of steady state tokamak reactor performance

    International Nuclear Information System (INIS)

    Devoto, R.S.; Fenstermacher, M.E.

    1990-01-01

    A new method of analysis for presenting the possible operating space for steady state, non-ignited tokamak reactors is proposed. The method uses contours of reactor performance and plasma characteristics, fusion power gain, wall neutron flux, current drive power, etc., plotted on a two-dimensional grid, the axes of which are the plasma current I p and the normalized beta, β n = β/(I p /aB 0 ), to show possible operating points. These steady state operating contour plots are called SOPCONS. This technique is illustrated in an application to a design for the International Thermonuclear Experimental Reactor (ITER) with neutral beam, lower hybrid and bootstrap current drive. The utility of the SOPCON plots for pointing out some of the non-intuitive considerations in steady state reactor design is shown. (author). Letter-to-the-editor. 16 refs, 3 figs, 1 tab

  5. Self operation type reactor scram device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1992-01-01

    A control rod having neutron absorbers therein is held by a curie point electromagnet by way of a control rod extension shaft. The electromagnet is suspended from a vertically movable driving shaft in an upper guide tube. Then, a heater is disposed at the lower portion in the inner side of the upper guide tube. Upon a function confirmation test, the electromagnet is at first pulled up to the inside of the upper guide tube. Subsequently, the electromagnet is heated by the heater by a temperature higher than the curie point of the temperature sensing magnetic material. If the function is normal, armature connected to the control rod extension tube is separated. With such a constitution, the electromagnetic portion is isolated from a coolant main stream, thereby enabling to avoid the cooling effect by the stream of coolants. Accordingly, the operation test for confirming the integrity of the function of the curie point electromagnet can be conducted while placing the electromagnet in the reactor core as it is during actual reactor operation. (I.N.)

  6. Self operation type reactor control device

    International Nuclear Information System (INIS)

    Saito, Makoto; Gunji, Minoru.

    1990-01-01

    A boiling-requefication chamber containing transporting materials having somewhat higher boiling point that the usual reactor operation temperature and liquid neutron absorbers having a boiling point sufficiently higher than that of the transporting materials is disposed near the coolant exit of a fuel assembly and connected with a tubular chamber in the reactor core with a moving pipe at the bottom. Since the transporting materials in the boiling-requefication chamber is boiled and expanded by heating, the liquid neutron absorbers are introduced passing through the moving pipe into the cylindrical chamber to control the nuclear reactions. When the temperature is lowered by the control, the transporting materials are liquefied to contract the volume and the liquid neutron absorbers in the cylindrical chamber are returned passing through the moving tube into the boiling-liquefication chamber to make the nuclear reaction vigorous. Thus, self-operation type power conditioning and power stopping are enabled not by way of control rods and not requiring external control, to prevent scram failure or misoperation. (N.H.)

  7. BWR type reactor and its operating method

    International Nuclear Information System (INIS)

    Ootsuji, Niro.

    1983-01-01

    Purpose: To regulate the control rod extraction operation such that an assumed control rod drop accident, if should occur, may not lead to further serious accidents, as well as enable to improve the working life of the control rod. Method: A plurality of control rods disposed among a plurality of fuel assemblies constituting the reactor core for suppressing the reactor core reactivity are divided into two groups depending on the descending speed, and the number of rods with a faster descending speed is set to less than 1/4 of the total number of the control rods. Then, the control rods are arranged such that those rods of the faster descending speed may be set every one another in any of the vertical, lateral and orthogonal directions. Further, it is always judged as to the possibility of extracting the control rods with the faster descending speed by a fast control rod extraction judging circuit to issue a signal to a control rod extraction inhibition circuit, so that the extraction operation for the control rods with the faster descending speed is started after all of the control rods with the slow descending speed have been extracted. Accordingly, if a control rod dropping accident should occur, abrupt power change can be avoided to thereby minimize the development of the accident. (Horiuchi, T.)

  8. ITER parametric analysis and operational performance

    International Nuclear Information System (INIS)

    Perkins, L.J.; Spears, W.R.; Galambos, J.D.

    1991-01-01

    One of the key components of the ITER Conceptual Design Activities (CDA) is the determination of optimum design, investigation of operation in various modes, recommendation of baseline performance specifications, studies of sensitivity of ITER design to uncertainties in physics, investigation of operational flexibility, assessment of alternative designs, and determination of implications for extrapolation to prospective DEMO reactors. These terms of reference are reported in this document. Refs, figs and tabs

  9. Regulation for installation and operation of marine reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The regulation is defined under the law for the regulations of nuclear source materials, nuclear fuel materials and reactors and the provisions of the order for execution of the law. The regulation is applied to marine reactors and reactors installed in foreign nuclear ships. Basic concepts and terms are explained, such as: radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; safeguarded area; inspected surrounding area and employee. The application for permission of installation of reactors shall list maximum continuous thermal power, location and general structure of reactor facilities, structure and equipment of reactors and treatment and storage facilities of nuclear fuel materials, etc. The application for permission of reactors installed in foreign ships shall describe specified matters according to the provisions for domestic reactors. The operation program of reactors for three years shall be filed to the Minister of Transportation for each reactor every fiscal year from that year when the operation is expected to start. Records shall be made for each reactor and kept for particular periods on inspection of reactor facilities, operation, fuel assembly, control of radiation, maintenance and others. Exposure doses, inspection and check up of reactor facilities, operation of reactors, transport and storage of nuclear fuel materials, etc. are designated in detail. (Okada, K.)

  10. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  11. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  12. 78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

    Science.gov (United States)

    2013-11-29

    ... NUCLEAR REGULATORY COMMISSION [NRC-2013-0260] Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence The U.S. Nuclear Regulatory Commission (NRC) is issuing this Federal Register notice to inform the public of a slight change in the manner of distribution of publicly available operating reactor licensing...

  13. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

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

  14. Current safety issues related to research reactor operation

    International Nuclear Information System (INIS)

    Alcala-Ruiz, F.

    2000-01-01

    The Agency has included activities on research reactor safety in its Programme and Budget (P and B) since its inception in 1957. Since then, these activities have traditionally been oriented to fulfil the Agency's functions and obligations. At the end of the decade of the eighties, the Agency's Research Reactor Safety Programme (RRSP) consisted of a limited number of tasks related to the preparation of safety related publications and the conduct of safety missions to research reactor facilities. It was at the beginning of the nineties when the RRSP was upgraded and expanded as a subprogramme of the Agency's P and B. This subprogramme continued including activities related to the above subjects and started addressing an increasing number of issues related to the current situation of research reactors (in operation and shut down) around the world such as reactor ageing, modifications and decommissioning. The present paper discusses some of the above issues as recognised by various external review or advisory groups (e.g., Peer Review Groups under the Agency's Performance Programme Appraisal System (PPAS) or the standing International Nuclear Safety Advisory Group (INSAG)) and the impact of their recommendations on the preparation and implementation of the part of the Agency's P and B relating to the above subject. (author)

  15. Operation of CANDU power reactor in thorium self-sufficient fuel cycle

    Indian Academy of Sciences (India)

    This paper presents the results of calculations for CANDU reactor operation in thorium fuel cycle. Calculations are performed to estimate the feasibility of operation of heavy-water thermal neutron power reactor in self-sufficient thorium cycle. Parameters of active core and scheme of fuel reloading were considered to be the ...

  16. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  17. Report of the reactor Operators Service - Annex F

    International Nuclear Information System (INIS)

    Zivotic, Z.

    1990-01-01

    RA reactor operators service is organized in two groups: permanent staff (chief operator, chief shift operators and operators) and changeable group which is formed according to the particular operation needs for working in shifts. During 1989 the operators service staff participated in the following activities: reconstruction of the existing reactor systems, control of the emergency cooling system, construction of the experimental loop 'Vinca-1'. Education of the staff was organized through routine courses, practical training is foreseen for 1991 [sr

  18. Five years of operating the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Benedict, Georg

    1970-01-01

    Considerable obstacles had to be surmounted before TRIGA MAINZ, first TRIGA reactor built in Germany, reached initial criticality in 1965. Subsequent five years' operation did not raise any major problems. The facility has proven quite reliable and particularly well suited for the purposes of the nuclear chemistry research program pursued at Mainz University. Extensive use is made of the pulse mode of operation. As a result, fuel elements are obviously somewhat overstressed, even though most pulses performed are of the 1.50 dollar size. Maximum licensed steady state power of 100 kW till now has met the requirements of most experiments. However, efforts are in progress to improve irradiation conditions by increasing the reactor power to 300 kW. (author)

  19. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    International Nuclear Information System (INIS)

    Lin Chaung; Shen Chihming

    2000-01-01

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller

  20. Operator reliability study for Probabilistic Safety Analysis of an operating research reactor

    International Nuclear Information System (INIS)

    Mohamed, F.; Hassan, A.; Yahaya, R.; Rahman, I.; Maskin, M.; Praktom, P.; Charlie, F.

    2015-01-01

    Highlights: • Human Reliability Analysis (HRA) for Level 1 Probabilistic Safety Analysis (PSA) is performed on research nuclear reactor. • Implemented qualitative HRA framework is addressed. • Human Failure Events of significant impact to the reactor safety are derived. - Abstract: A Level 1 Probabilistic Safety Analysis (PSA) for the TRIGA Mark II research reactor of Malaysian Nuclear Agency has been developed to evaluate the potential risk in its operation. In conjunction to this PSA development, Human Reliability Analysis (HRA) is performed in order to determine human contribution to the risk. The aim of this study is to qualitatively analyze human actions (HAs) involved in the operation of this reactor according to the qualitative part of the HRA framework for PSA which is namely the identification, qualitative screening and modeling of HAs. By performing this framework, Human Failure Events (HFEs) of significant impact to the reactor safety are systematically analyzed and incorporated into the PSA structure. A part of the findings in this study will become the input for the subsequent quantitative part of the HRA framework, i.e. the Human Error Probability (HEP) quantification

  1. The development of reactor operator license examination question bank

    International Nuclear Information System (INIS)

    Kim, In Hwan; Woo, S. M.; Kam, S. C.; Nam, K. J.; Lim, H. P.

    2001-12-01

    The number of NPP keeps increasing therefore there is more need of reactor operators. This trend requires the more efficiency in managing the license examination. Question bank system will help us to develop good quality examination materials and keep them in it. The ultimate purpose of the bank system is for selecting qualified reactor operators who are primarily responsible for the safety of reactor operation in NPP

  2. RA reactor operation and maintenance in 1996, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1996-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. Since the RA reactor is shutdown since 1984, it is high time for decision making of its future status. Possible solutions for the future status of the RA reactor discussed in this report are: renewal of reactor components for the reactor restart, conservation of the reactor (temporary shutdown) or permanent reactor shutdown. Control and maintenance of the reactor instrumentation and devices was done regularly but dependent on the availability of the spare parts and financial means. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  3. RA Reactor operation and maintenance (I-IX), Part I

    International Nuclear Information System (INIS)

    Zecevic, V.

    1963-12-01

    The report on RA reactor operation and maintenance for year 1963 is divided in six tasks. This volume contains the introductory report, and three tasks of the final report, namely reactor exploitation, reactivity changes of the RA reactor before repair, planning of refuelling

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

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Akira; Takei, Hiroaki; Iwanaga, Shigeru; Noda, Masao; Hara, Hidemi (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

    1983-09-01

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

  5. A series of lectures on operational physics of power reactors

    International Nuclear Information System (INIS)

    Mohanakrishnan, P.; Rastogi, B.P.

    1982-01-01

    This report discusses certain aspects of operational physics of power reactors. These form a lecture series at the Winter College on Nuclear Physics and Reactors, Jan. - March 1980, conducted at the International Centre for Theoretical Physics, Trieste, Italy. The topics covered are (a) the reactor physics aspects of fuel burnup (b) theoretical methods applied for burnup prediction in power reactors (c) interpretation of neutron detector readings in terms of adjacent fuel assembly powers (d) refuelling schemes used in power reactors. The reactor types chosen for the discussion are BWR, PWR and PHWR. (author)

  6. Regulation for installation and operation of experimental-research reactor

    International Nuclear Information System (INIS)

    1979-01-01

    The ordinance is stipulated under the Law for regulation of nuclear raw materials, nuclear fuel materials and reactors and the provisions for installation and operation of reactor in the order for execution of the law. Basic concepts and terms are defined, such as, radioactive waste; fuel assembly; exposure dose; accumulative dose; controlled area; preserved area; inspected surrounding area and employee. An application for permission of installation of reactor shall list such matters as: the maximum continuous thermal output of reactor; location and general construction of reactor facilities; construction and equipment of the main reactor and other facilities for nuclear fuel materials; cooling and controlling system and radioactive waste, etc. An operation plan of reactor for three years shall be filed till January 31 of the fiscal year preceding that one the operation begins. Records shall be made and kept for specified periods respectively on inspection of reactor facilities, operation, fuel assembly, radiation control, maintenance, accidents of reactor equipment and weather. Detailed rules are settled for entrance limitation to controlled area, exposure dose, inspection, check up and regular independent examination of reactor facilities, operation of reactor, transportation of substances contaminated by nuclear fuel materials within the works and storage, etc. (Okada, K.)

  7. Neutronic Core Performance of CAREM-25 Reactor

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel; Matzkin, S

    2000-01-01

    The actual design state of core of CAREM-25 reactor is presented.It is shown that the core design complains with the safety and operation established requirements.It is analyzed the behavior of the reactor safety and control systems (single failure of the fast shut down system, single failure of the shut down system, single failure of the second shut down system, reactivity worth of the adjust and control system in normal operation and hot shut down, reactivity worth of the adjust and control system and the scheme of movement of the control rod during the operation cycle).It is shown the burnup profile of fuel elements with the proposed scheme of refueling and the burnup and power density distribution at different moments of the operation cycle.The power peaking factor of the equilibrium core is 2.56, the minimum DNBR is 1.90 and its average is 2.09 during the operation cycle

  8. Effect of beta limits on reactor performance in EBT

    International Nuclear Information System (INIS)

    Uckan, N.A.; Spong, D.A.; Nelson, D.B.

    1981-01-01

    Because of uncertainties in extrapolating results of simplified models to a reactor plasma, the parameters that influence the beta limits cannot be determined accurately at the present time. Also, the reasonable changes within the models and/or assumptions are seen to affect the core beta limits by almost an order of magnitde. Hence, at the present, these limits cannot be used as a rigid (and reliable) requirement for ELMO Bumpy Torus (EBT) reactor engineering considerations. However, sensitivity studies can be carried out to determine the boundaries of the operating regime and to demonstrate the effects of various modes, assumptions, and models on reactor performance (Q value). First, the modes believed to limit the core β and ring plasma performance are discussed, and the simplifications and/or assumptions involved in deriving these limits are highlighted. Then, the implications of these limits for a reactor are given

  9. Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

    Science.gov (United States)

    Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

    2018-04-01

    1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

  10. Design and Construction of Operation Bridge for Research Reactor

    International Nuclear Information System (INIS)

    Jung, Kwangsub; Choi, Jinbok; Lee, Jongmin; Oh, Jinho

    2015-01-01

    The operation bridge contains a lower working deck mounted on a saddle that travels on rails. Upright members are mounted on the saddle to support the upper structure and two hoist monorails. The saddle contains an anti-derail system that is composed of seismic lugs and guide rollers. The operation bridge travels along the rails to transport the fuel assembly, irradiated object, and reactor components in the pools by using tools. Hoists are installed at the top girder. The hoist is suspended from the monorail by means of a motor driven trolley that runs along the monorail. Movements of hoist and trolley are controlled by using the control pendant switch. Processes of design and construction of the operation bridge for the research reactor are introduced. The operation bridge is designed under consideration of functions of handling equipment in the pool and operational limits for safety. Structural analysis is carried out to evaluate the structural integrity in the seismic events. Tests and inspections are also performed during fabrication and installation to confirm the function and safety of the operation bridge

  11. Performance improvement of the Annular Core Pulse Reactor for reactor safety experiments

    International Nuclear Information System (INIS)

    Reuscher, J.A.; Pickard, P.S.

    1976-01-01

    The Annular Core Pulse Reactor (ACPR) is a TRIGA type reactor which has been in operation at Sandia Laboratories since 1967. The reactor is utilized in a wide variety of experimental programs which include radiation effects, neutron radiography, activation analysis, and fast reactor safety. During the past several years, the ACPR has become an important experimental facility for the United States Fast Reactor Safety Research Program and questions of interest to the safety of the LMFBR are being addressed. In order to enhance the capabilities of the ACPR for reactor safety experiments, a project to improve the performance of the reactor was initiated. It is anticipated that the pulse fluence can be increased by a factor of 2.0 to 2.5 utilizing a two-region core concept with high heat capacity fuel elements around the central irradiation cavity. In addition, the steady-state power of the reactor will be increased by about a factor of two. The new features of the improvements are described

  12. Operational human performance reliability assessment (OHPRA)

    International Nuclear Information System (INIS)

    Haas, P.M.; Swanson, P.J.; Connelly, E.M.

    1993-01-01

    Operational Human Performance Reliability Assessment (OHPRA) is an approach for assessing human performance that is being developed in response to demands from modern process industries for practical and effective tools to assess and improve human performance, and therefore overall system performance and safety. The single most distinguishing feature of the approach is that is defines human performance in open-quotes operationalclose quotes terms. OHPRA is focused not on generation of human error probabilities, but on practical analysis of human performance to aid management in (1) identifying open-quotes fixableclose quotes problems and (2) providing input on the importance and nature of potential improvements. Development of the model in progress uses a unique approach for eliciting expert strategies for assessing performance. A PC-based model incorporating this expertise is planned. A preliminary version of the approach has already been used successfully to identify practical human performance problems in reactor and chemical process plant operations

  13. Core design concepts for high performance light water reactors

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.

    2007-01-01

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

  14. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  15. Reactor performances and microbial communities of biogas reactors: effects of inoculum sources.

    Science.gov (United States)

    Han, Sheng; Liu, Yafeng; Zhang, Shicheng; Luo, Gang

    2016-01-01

    Anaerobic digestion is a very complex process that is mediated by various microorganisms, and the understanding of the microbial community assembly and its corresponding function is critical in order to better control the anaerobic process. The present study investigated the effect of different inocula on the microbial community assembly in biogas reactors treating cellulose with various inocula, and three parallel biogas reactors with the same inoculum were also operated in order to reveal the reproducibility of both microbial communities and functions of the biogas reactors. The results showed that the biogas production, volatile fatty acid (VFA) concentrations, and pH were different for the biogas reactors with different inocula, and different steady-state microbial community patterns were also obtained in different biogas reactors as reflected by Bray-Curtis similarity matrices and taxonomic classification. It indicated that inoculum played an important role in shaping the microbial communities of biogas reactor in the present study, and the microbial community assembly in biogas reactor did not follow the niche-based ecology theory. Furthermore, it was found that the microbial communities and reactor performances of parallel biogas reactors with the same inoculum were different, which could be explained by the neutral-based ecology theory and stochastic factors should played important roles in the microbial community assembly in the biogas reactors. The Bray-Curtis similarity matrices analysis suggested that inoculum affected more on the microbial community assembly compared to stochastic factors, since the samples with different inocula had lower similarity (10-20 %) compared to the samples from the parallel biogas reactors (30 %).

  16. Supervisory and managerial aspects on the safe operation of the Egyptian second research reactor

    International Nuclear Information System (INIS)

    Abdelrazek, I.D.; Shokr, A.M.

    2000-01-01

    ETRR-2 is a multipurpose reactor for radioisotopes production, neutron beam experiments, basic and applied research in physics and engineering, fuels and material tests, and for training. The reactor is an open pool type, 22 MW power, with average thermal neutron flux of 1.4x10 14 n/cm 2 sec, cooled and moderated by light water and with beryllium reflectors. Various experimental devices and irradiation facilities are integrated with the reactor. The reactor has been licensed for operation in November 1998. Several principles and regulations have been applied to all the reactor project stages to achieve safety. Moreover, other several principals, regulations, and aspects are enforced by the AEA, National Center for Nuclear Safety and Radiation Control, NCNSRC, and reactor management to achieve safety during reactor operation and utilization. Responsibility on Safety and Supervision Aspects AEA chairman has the ultimate responsibility on the reactor safety during operation and utilization. The primary responsible on the safety is the ETRR-2 supervisor, who is supervising the ETRR-2 site that includes the ETRR-2 reactor, Fuel Manufacturing Plant, and other two projects (under execution): Radioisotopes Production Plant and Dry Fuel Storage Facility. ETRR-2 supervisor is responsible to ensure that: the reactor is operated in accordance with the safety requirements by qualified and trained personnel, updating and enforcement of the reactor mandatory documentation, and the services are adequate for the reactor operation. He is responsible, also, to guarantee that: the reactor manager has enough authority and resources to carry out his function effectively and the reactor is kept operated in agreement with established procedures. A Technical Revision Committee, TRC, is formed to advice the ETRR-2 supervisor on the safety of the reactor and experiments. The committee members are from AEA experts with no direct relation to the reactor and experiments being performed. Members

  17. Operating reactors licensing actions summary. Volume 5, No. 6

    International Nuclear Information System (INIS)

    1985-08-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published for internal NRC use in managing the Operating Reactors Licensing Actions Program. Its content will change based on NRC management informational requirements

  18. Operating reactors licensing actions summary. Vol. 3, No. 6

    International Nuclear Information System (INIS)

    1983-07-01

    The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

  19. Sustainable operations in nuclear research reactors. A bibliographical study

    International Nuclear Information System (INIS)

    Kibrit, Eduardo; Rodrigues de Aquino, Afonso; Marotti de Mello, Adriana; Tromboni de Souza Nascimento, Paulo

    2017-01-01

    Sustainability is gaining prominence in the area of operations management. By means of a bibliographical research, we identified in literature sustainable operations carried out by operating organizations of nuclear research reactors. The methodology applied consisted in gathering material, descriptive analysis, selection of analytical categories and evaluation of the material collected. The collection of material was performed by a search made on academic and nuclear databases, with keywords structured for the subject of the research. The collected material was analysed and analytical categories on the theme sustainable operations were established. The evaluation of the collected material resulted in references accepted for the study, classified according to the pre-established analytical categories. The results were significant. From then on, a theoretical review on the topic under study was structured, based on pre-defined analytical categories. Thus, we were able to identify gaps in the literature and propose new studies on the subject.

  20. Sustainable operations in nuclear research reactors. A bibliographical study

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo; Rodrigues de Aquino, Afonso [Cidade Univ., Sao Paolo (Brazil). Inst. de Pesquisas Energeticas e Nucleares; Marotti de Mello, Adriana [Sao Paolo Univ. (Brazil). Faculdade de Economia; Tromboni de Souza Nascimento, Paulo [Sao Paolo Univ. (Brazil). Faculdade de Economia Administracao e Contabilidade

    2017-10-15

    Sustainability is gaining prominence in the area of operations management. By means of a bibliographical research, we identified in literature sustainable operations carried out by operating organizations of nuclear research reactors. The methodology applied consisted in gathering material, descriptive analysis, selection of analytical categories and evaluation of the material collected. The collection of material was performed by a search made on academic and nuclear databases, with keywords structured for the subject of the research. The collected material was analysed and analytical categories on the theme sustainable operations were established. The evaluation of the collected material resulted in references accepted for the study, classified according to the pre-established analytical categories. The results were significant. From then on, a theoretical review on the topic under study was structured, based on pre-defined analytical categories. Thus, we were able to identify gaps in the literature and propose new studies on the subject.

  1. INDIAN POINT REACTOR STARTUP AND PERFORMANCE

    Energy Technology Data Exchange (ETDEWEB)

    Deddens, J. C.; Batch, M. L.

    1963-09-15

    The testing program for the Indian Point Reactor is discussed. The thermal and hydraulic evaluation of the primary coolant system is discussed. Analyses of fuel loading and initial criticality, measurement of operating coefficients of reactivity, control rod group reactivity worths, and xenon evaluation are presented. (R.E.U.)

  2. Operating reactors licensing actions summary. Volume 5, Number 1

    International Nuclear Information System (INIS)

    1985-03-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

  3. General areas needing chemical competence to support reactor operation

    International Nuclear Information System (INIS)

    Proksch, E.; Bildstein, H.

    1963-01-01

    Chemical competence is needed not only for the development of new types of reactors but also for the start-up and safe operation of reactors. The activities of chemistry and chemical engineering cover a number of fields, namely chemical analysis, radiochemical analysis, corrosion research, radiolysis of water and water purification. The author reviews fields in reactor operation and maintenance in which chemical competence is needed. (author). 9 refs

  4. Advanced CANDU reactor design for operability

    International Nuclear Information System (INIS)

    Hopwood, J.M.; Lalonde, R.; Soulard, M.

    2003-01-01

    This paper outlines design features and engineering processes in the ACR TM development program which contribute to excellence in performance and low operating cost. AECL recognizes that future plant owners will place a high priority in these operational characteristics. A successful next generation plant will have a best-in-class capability, both in its design characteristics, in the engineering philosophy and program adopted during the product development, and in the vendor's approach to operating station support. The ACR program addresses each of these drivers. Operability considerations are built-in to the design at an overall, plant wide level. For example, based on the strong CANDU 6 operating record, targets for standard outage duration, time between outages and component durability are set, while the design engineering is managed to achieve these targets. The ultimate maintenance target for the ACR, once initial operating experience has been gained, is to operate with a 21-day standard maintenance outage at an interval of once every three years. At the detailed design level, close attention is paid to space allocation, to enable good maintenance access. Selection of components also places emphasis on maintainability based on the extensive and current experience with CANDU projects. (author)

  5. International Experience with Fast Reactor Operation & Testing

    International Nuclear Information System (INIS)

    Sackett, John I.; Grandy, C.

    2013-01-01

    Conclusion: • Worldwide experience with fast reactors has demonstrated the robustness of the technology and it stands ready for worldwide deployment. • The lessons learned are many and there is danger that what has been learned will be forgotten given that there is little activity in fast reactor development at the present time. • For this reason it is essential that knowledge of fast reactor technology be preserved, an activity supported in the U.S. as well as other countries

  6. Neutron field control cybernetics model of RBMK reactor operator

    International Nuclear Information System (INIS)

    Polyakov, V.V.; Postnikov, V.V.; Sviridenkov, A.N.

    1992-01-01

    Results on parameter optimization for cybernetics model of RBMK reactor operator by power release control function are presented. Convolutions of various criteria applied previously in algorithms of the program 'Adviser to reactor operator' formed the basis of the model. 7 refs.; 4 figs

  7. Strategies of operation cycles in BWR type reactors

    International Nuclear Information System (INIS)

    Molina, D.; Sendino, F.

    1996-01-01

    The article analyzes the operation cycles in BWR type reactors. The cycle size of operation is the consequence on the optimization process of the costs with the technical characteristics of nuclear fuel and the characteristics of demand and production. The authors analyze the cases of Garona NP and Cofrentes NP, both with BWR reactors. (Author)

  8. High-intensity power-resolved radiation imaging of an operational nuclear reactor.

    Science.gov (United States)

    Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

    2015-10-09

    Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

  9. Multilayer robust control for safety enhancement of reactor operations

    International Nuclear Information System (INIS)

    Edwards, R.M.; Lee, K.Y.; Ray, A.

    1991-01-01

    A novel concept of reactor power and temperature control has been recently reported in which a conventional output feedback controller is embedded within a state feedback setting. The embedded output feedback controller at the inner layer largely compensates for plant modeling uncertainties and external disturbances, and the outer layer generates an optimal control signal via feedback of the estimated plant states. A major advantage of this embedded architecture is the robustness of the control system relative to parametric and nonparametric uncertainties and thus the opportunity for designing fault-accommodating control algorithms to improve reactor operations and plant safety. The paper illustrates the architecture of the state-feedback-assisted classical (SFAC) control, which utilizes an embedded output feedback controller designed via classical techniques. It demonstrates the difference between the performance of conventional state feedback control and SFAC by examining the sensitivity of the dominant eigenvalues of the individual closed-loop systems

  10. Proposed Reactor Operating Experience Feedback System Development

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Kim, Min Chul; Huh, Chang Wook; Lee, Durk Hun; Bae, Koo Hyun

    2006-01-01

    Most events occurring in nuclear power plants are not individually significant, and prevented from progressing to accident conditions by a series of barriers against core damage and radioactive releases. Significant events, if occur, are almost always a breach of these multiple barriers. As illustrated in the 'Swiss cheese' model, the individual layers of defense or 'cheese slices' have weakness or 'holes.' These weaknesses are inconstant, i.e., the holes are open or close at random. When by chance all the holes are aligned, a hazard causes the significant event of concern. Elements of low significant events, inattention to detail, time or economic pressure, uncorrected poor practices/habits, marginal maintenance and equipment care, etc., make holes in the layers of defense; some elements may make more holes in different layers, incurring more chances to be aligned. An effective reduction of the holes, therefore, is gained through better knowledge or awareness of increasing trends of the event elements, followed by appropriate actions. According to the Swiss cheese metaphor, attention to the Operating Experience (OE) feedback system, as opposed to the individual and to randomness, is drawn from a viewpoint of reactor safety

  11. Annual report on operation, utilization and technical development of research reactors and hot laboratory

    International Nuclear Information System (INIS)

    1990-09-01

    This report describes the activities of the Department of Research Reactor Operation in fiscal year of 1989. It also presents some technical topics on the reactor operation and utilization in details. The Department is responsible for operation of the research reactors, JRR-2 and JRR-4, and the Hot Laboratory. The research reactor JRR-3 was reconstructed to enhance the performance for utilization. The first criticality was achieved on March 22, 1989, and it subsequently went into operation. In connection with the reactor operation, the various research and development activities in the area of fuel management, water chemistry, radiation monitoring and material irradiation have been made. In the Hot Laboratory, post-irradiation examinations of fuels and materials have been carried out along with the development of related techniques. (author)

  12. The heating operational summarization in three winters of a 5 MW test heating reactor

    International Nuclear Information System (INIS)

    Wang Dazhong; Dong Duo; Su Qingshan; Zhang Yajun

    1992-09-01

    The 5 MW THR (5 MW test heating reactor) is a new type reactor with inherent safety developed by INET (Institute of Nuclear Energy Technology). It is the first 'pressure vessel type' heating reactor in operation in the world. It was put into operation in November, 1989. Since then it has operated for three winter seasons. The total operation time has reached to 8174 hours and its availability of heating has reached to 99%. The advanced technology of this reactor has been proved in the past three years operation. The characteristics of power regulating, load following, reactivity disturbance and the variation of parameters under the condition of ATWS (anticipated transients without scram) were studied with experiments in 5 MW THR. The 5 MW THR is an ideal heating reactor and has outstanding performances

  13. Comparison of Ontario Hydro's performance with world power reactors - 1981

    International Nuclear Information System (INIS)

    Dumka, B.R.

    1982-04-01

    The performance of Ontario Hydro's CANDU reactors in 1981 is compared with that of 123 world nuclear power reactors rated at 500 MW(e) or greater. The report is based on data extracted from publications, as well as correspondence with a number of utilities. The basis used is the gross capacity factor, which is defined as gross unit generation divided by the perfect gross output for the period of interest. The lowest of the published turbine and generator design ratings is used to determine the perfect gross output, unless the unit has been proven capable of consistently exceeding this value. The first six reactors in the rankings were CANDU reactors operated by Ontario Hydro

  14. Operational experience with Dragon reactor experiment of relevance to commercial reactors

    International Nuclear Information System (INIS)

    Capp, P.D.; Simon, R.A.

    1976-01-01

    An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

  15. RA reactor operation and maintenance in 1994, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Cupac, S.; Sulem, B.; Zivotic, Z.; Mikic, N.; Tanaskovic, M.

    1994-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The planned major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the reactor power supply system. The existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Renewal of the reactor instrumentation was started but but it is behind the schedule because the delivery of components from USSR was stopped for political reasons. The spent fuel elements used from the very beginning of reactor operation are stored in the existing pools. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988 and was fulfilled in 1990. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  16. OPAD: An expert system for research reactor operations and fault diagnosis using probabilistic safety assessment tools

    International Nuclear Information System (INIS)

    Verma, A.K.; Varde, P.V.; Sankar, S.; Prakash, P.

    1996-01-01

    A prototype Knowledge Based (KB) operator Adviser (OPAD) system has been developed for 100 MW(th) Heavy Water moderated, cooled and Natural Uranium fueled research reactor. The development objective of this system is to improve reliability of operator action and hence the reactor safety at the time of crises as well as normal operation. The jobs performed by this system include alarm analysis, transient identification, reactor safety status monitoring, qualitative fault diagnosis and procedure generation in reactor operation. In order to address safety objectives at various stages of the Operator Adviser (OPAD) system development the Knowledge has been structured using PSA tools/information in an shell environment. To demonstrate the feasibility of using a combination of KB approach with PSA for operator adviser system, salient features of some of the important modules (viz. FUELEX, LOOPEX and LOCAEX) have been discussed. It has been found that this system can serve as an efficient operator support system

  17. Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

    1974-07-01

    Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

  18. Operation and maintenance of the RA Reactor in 1985, Part 1, Annex A - Reactor applications

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

    This document describes reactor operation from 1981 to 1985, including data about short term (shorter than 24 hours) and long term operation interruptions, as well as safety shutdown and reactor applications. During 1982, 1983 until July 1984 reactor was operated at 2 MW power according to the plan. Plan was not fulfilled in 1983 because deposits were noticed again, at the end of 1982, on the surface of fuel elements. Reactor was mainly used for neutron activation purposes and isotope production as source of neutrons for experimental purposes [sr

  19. Radiological monitoring related to the operation of PUSPATI's Triga Reactor

    International Nuclear Information System (INIS)

    Fatimah Mohamad Amin; Mohamad Yusof Mohamad Ali; Lau How Mooi; Idris Besar.

    1983-01-01

    Reactor operation is one of the main activities carried out at the Tun Ismail Atomic Research Centre (PUSPATI) which requires radiological monitoring. This paper describes the programme for radiological monitoring which is related to the operation of the 1 MW Triga MK II research reactor which was commissioned in July, 1982. This programme includes monitoring of the radiation and contamination levels of the reactor and its associated facilities and environmental monitoring of PUSPATI's site and its environs. The data presented in this paper covers the period between 1982 to 1983 which includes both the pre-operational and operational phases of the monitoring programme. (author)

  20. Annual report 1989 operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1989-01-01

    In 1989 the operation of the High Flux Reactor Petten was carried out as planned. The availability was more than 100% of scheduled operating time. The average occupation of the reactor by experimental devices was 72% of the practical occupation limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons and for radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  1. Annual report 1990. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1990-01-01

    In 1990 the operation of the High Flux Reactor was carried out as planned. The availability was 96% of scheduled operating time. The average utilization of the reactor was 71% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  2. Annual Report 1991. Operation of the high flux reactor

    International Nuclear Information System (INIS)

    Ahlf, J.; Gevers, A.

    1992-01-01

    In 1991 the operation of the High Flux Reactor was carried out as planned. The availability was more than 100% of scheduled operating time. The average utilization of the reactor was 69% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. Development activities addressed upgrading of irradiation devices, neutron capture therapy, neutron radiography and neutron transmutation doping of silicon. General activities in support of running irradiation programmes progressed in the normal way

  3. Operating experience with the DRAGON High Temperature Reactor experiment

    International Nuclear Information System (INIS)

    Simon, R.A.; Capp, P.D.

    2002-01-01

    The Dragon Reactor Experiment in Winfrith/UK was a materials test facility for a number of HTR projects pursued in the sixties and seventies of the last century. It was built and managed as an OECD/NEA international joint undertaking. The reactor operated successfully between 1964 and 1975 to satisfy the growing demand for irradiation testing of fuels and fuel elements as well as for technological tests of components and materials. The paper describes the reactor's main experimental features and presents results of 11 years of reactor operation relevant for future HTRs. (author)

  4. Effect of post-digestion temperature on serial CSTR biogas reactor performance

    DEFF Research Database (Denmark)

    Boe, Kanokwan; Karakashev, Dimitar Borisov; Trably, Eric

    2009-01-01

    The effect of post-digestion temperature on a lab-scale serial continuous-flow stirred tank reactor (CSTR) system performance was investigated. The system consisted of a main reactor operated at 55 degrees C with hydraulic retention time (HRT) of 15 days followed by post-digestion reactors with HRT...

  5. From USA operation experience of industrial uranium-graphite reactors

    International Nuclear Information System (INIS)

    Burdakov, N.S.

    1996-01-01

    The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors

  6. Massive computation methodology for reactor operation (MACRO)

    International Nuclear Information System (INIS)

    Gustavsson, Cecilia; Pomp, Stephan; Sjoestrand, Henrik; Wallin, Gustav; Oesterlund, Michael; Koning, Arjan; Rochman, Dimitri; Bejmer, Klaes-Hakan; Henriksson, Hans

    2010-01-01

    Today, nuclear data libraries do not handle uncertainties from nuclear data in a consistent manner and the reactor codes do not request uncertainties in nuclear data input. Thus, the output from these codes have unknown uncertainties. The plan is to use a method proposed by Koning and Rochman to investigate the propagation of nuclear data uncertainties into reactor physics codes and macroscopic parameters. A project (acronym MACRO) has started at Uppsala University in collaboration with A. Koning and with financial support from Vattenfall AB and the Swedish Research Council within the GENIUS (Generation IV research in universities of Sweden) project. In the proposed method the uncertainties in nuclear model parameters will be derived from theoretical considerations and comparisons of nuclear model results with experimental cross-section data. Given the probability distribution in the model parameters a large set of random, complete ENDF-formatted nuclear data libraries will be created using the TALYS code. The generated nuclear data libraries will then be used in neutron transport codes to obtain macroscopic reactor parameters. For this, models of reactor systems with proper geometry and elements will be used. This will be done for all data libraries and the variation of the final results will be regarded as a systematic uncertainty in the investigated reactor parameter. The understanding of these systematic uncertainties is especially important for the design and intercomparison of new reactor concepts, i.e., Generation IV, and optimization applications for current generation reactors is envisaged. (authors)

  7. Massive computation methodology for reactor operation (MACRO)

    Energy Technology Data Exchange (ETDEWEB)

    Gustavsson, Cecilia; Pomp, Stephan; Sjoestrand, Henrik; Wallin, Gustav; Oesterlund, Michael [Division of applied nuclear physics, Department of physics and astronomy, Uppsala University, Laegerhyddsvaegen 1, 751 20 Uppsala (Sweden); Koning, Arjan; Rochman, Dimitri [Nuclear Research and consultancy Group (NRG) Westerduinweg 3, Petten (Netherlands); Bejmer, Klaes-Hakan [Vattenfall Nuclear Fuel AB, Jaemtlandsgatan 99, Vaellingby (Sweden); Henriksson, Hans [Vattenfall Research and Development AB, Jaemtlandsgatan 99, Vaellingby (Sweden)

    2010-07-01

    Today, nuclear data libraries do not handle uncertainties from nuclear data in a consistent manner and the reactor codes do not request uncertainties in nuclear data input. Thus, the output from these codes have unknown uncertainties. The plan is to use a method proposed by Koning and Rochman to investigate the propagation of nuclear data uncertainties into reactor physics codes and macroscopic parameters. A project (acronym MACRO) has started at Uppsala University in collaboration with A. Koning and with financial support from Vattenfall AB and the Swedish Research Council within the GENIUS (Generation IV research in universities of Sweden) project. In the proposed method the uncertainties in nuclear model parameters will be derived from theoretical considerations and comparisons of nuclear model results with experimental cross-section data. Given the probability distribution in the model parameters a large set of random, complete ENDF-formatted nuclear data libraries will be created using the TALYS code. The generated nuclear data libraries will then be used in neutron transport codes to obtain macroscopic reactor parameters. For this, models of reactor systems with proper geometry and elements will be used. This will be done for all data libraries and the variation of the final results will be regarded as a systematic uncertainty in the investigated reactor parameter. The understanding of these systematic uncertainties is especially important for the design and intercomparison of new reactor concepts, i.e., Generation IV, and optimization applications for current generation reactors is envisaged. (authors)

  8. RB Research nuclear reactor, 30 years of operation; Istrazivacki nuclearni reaktor RB, povodom 30 godina rada

    Energy Technology Data Exchange (ETDEWEB)

    Pesic, M; Stefanovic, D [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1988-06-15

    Paper describes utilization, modifications and changes of construction and control-safety systems done at the RB reactor during 30 years of operation. Experiments performed at the reactor are summarized, new reactor equipment is described and the future plans are shown. Rad prikazuje eksploataciju reaktora RB tokom 30 godina rada, modifikacije i izmene u konstruktivnim i upravljacko-sigurnosnim sistemima. Sumirani su eksperimenti izvedeni na njemu, prikazana je nova oprema i planovi za buduci rad.

  9. Knowledge-Based operation planning system for boiling water reactors

    International Nuclear Information System (INIS)

    Tatsuya Iwamoto; Shungo Sakurai; Hitoshi Uematsu; Makoto Tsuiki

    1987-01-01

    A knowledge-Based Boiling Water Reactor operation planning system was developed to support core operators or core management engineers in making core operation plans, by automatically generating suboptimum core operation procedures. The procedures are obtained by searching a branching tree of the possible core status (nodes) and the elementary operations to change the core status (branches). A path that ends at the target node, and contains only operationally feasible nodes can be a candidate of the solution. The core eigenvalue, the power distribution and the thermal limit parameters at key points are calculated by running a three-dimensional (3-D) BWR core physics simulator to examine the feasibility of the nodes and the performance of candidates. To obtain a practically acceptable solution within a reasonable time rather than making a time-consuming effort to get the optimum one, the Depth-First-Search method, together with the heuristic branch-bounding, was used to search the branching tree. The system was applied to actual operation plannings with real plant data, and gave satisfactory results. It can be concluded that the system can be applied to generate core operation procedures as a substitute for core management experts

  10. Small nuclear reactor safety design requirements for autonomous operation

    International Nuclear Information System (INIS)

    Kozier, K.S.; Kupca, S.

    1997-01-01

    Small nuclear power reactors offer compelling safety advantages in terms of the limited consequences that can arise from major accident events and the enhanced ability to use reliable, passive means to eliminate their occurrence by design. Accordingly, for some small reactor designs featuring a high degree of safety autonomy, it may be-possible to delineate a ''safety envelope'' for a given set of reactor circumstances within which safe reactor operation can be guaranteed without outside intervention for time periods of practical significance (i.e., days or weeks). The capability to operate a small reactor without the need for highly skilled technical staff permanently present, but with continuous remote monitoring, would aid the economic case for small reactors, simplify their use in remote regions and enhance safety by limiting the potential for accidents initiated by inappropriate operator action. This paper considers some of the technical design options and issues associated with the use of small power reactors in an autonomous mode for limited periods. The focus is on systems that are suitable for a variety of applications, producing steam for electricity generation, district heating, water desalination and/or marine propulsion. Near-term prospects at low power levels favour the use of pressurized, light-water-cooled reactor designs, among which those having an integral core arrangement appear to offer cost and passive-safety advantages. Small integral pressurized water reactors have been studied in many countries, including the test operation of prototype systems. (author)

  11. Operating manual for the Health Physics Research Reactor

    International Nuclear Information System (INIS)

    1985-11-01

    This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs

  12. Efficient compliance with prescribed bounds on operational parameters by means of hypothesis testing using reactor data

    International Nuclear Information System (INIS)

    Sermer, P.; Olive, C.; Hoppe, F.M.

    2000-01-01

    - A common problem in any reactor operations is to comply with a requirement that certain operational parameters are constrained to lie within some prescribed bounds. The fundamental issue which is to be addressed in any compliance description can be stated as follows: The compliance definition, compliance procedures and allowances for uncertainties in data and accompanying methodologies, should be well defined and justifiable. To this end, a mathematical framework for compliance, in which the computed or measured estimates of process parameters are considered random variables, is described in this paper. This allows a statistical formulation of the definition of compliance with licence or otherwise imposed limits. An important aspect of the proposed methodology is that the derived statistical tests are obtained by a Monte Carlo procedure using actual reactor operational data. The implementation of the methodology requires a routine surveillance of the reactor core in order to perform the underlying statistical tests. The additional work required for surveillance is balanced by the fact that the resulting actions on the reactor operations, implemented in station procedures, make the reactor 'safer' by increasing the operating margins. Furthermore, increased margins are also achieved by efficient solution techniques which may allow an increase in reactor power. A rigorous analysis of a compliance problem using statistical hypothesis testing based on extreme value probability distributions and actual reactor operational data leads to effective solutions in the areas of licensing, nuclear safety, reliability and competitiveness of operating nuclear reactors. (author)

  13. Application of noise analysis on the operation of the RSG-GAS reactor

    International Nuclear Information System (INIS)

    Surian P, Tukiran S

    1999-01-01

    The RSG-GAS reactor has been operating for radioisotope production and experiments so that it is necessary to perform analysis of the reactor operation. analysis was done based on reactor noise experiment. Neutron noise at low and high power of the RSG-GAS has been analyzed using time and frequency domain with aim to determine the safety of reactor operation. The safety of reactor operation based on two parameters as, prompt neutron decay constant and decay ratio. The parameters are useful for reactor operation, so it is necessary to determine accurately. For determining prompt neutron decay constant, neutron density in the core of reactor which operated at 10 k W, was collected by using Fission Chamber detectors (FC). Based on power spectral density (PSD) was achieved break frequency about 23 Hz, so that the prompt neutron decay constant is about 151 sec -1 . While at at high power 20 MW, neutron density was collected by using Compensated Ionization Chamber (CIC) detector. The result at high power showed that there is reactivity effect in the core because of fluctuation in temperature and density of the coolant, and the decay ratio of 0.20, showed that the reactor is still operation in stable

  14. Performance of nuclear fuel in the Krsko reactor; Spremljanje delovanja jedrskega goriva v reaktorju NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Jurcevic, M; Kurincic, B; Levstek, M F; Sambo, B; Vrcko, P [Nuklearna elektrana Krsko, Krsko (Yugoslavia)

    1987-07-01

    In this paper activities to follow performance of the nuclear fuel and operational status of the reactor of Nuclear Power Plant Krsko are presented. Short descriptions of the methods as well as nuclear and process instrumentation used for surveillance of the reactor performance are given. The purpose of the subject activities is to assure safe operation of the reactor in accordance with the Final safety Analysis Report of NPP Krsko. (author)

  15. RA Research nuclear reactor Part 1, RA Reactor operation and maintenance in 1987

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1987-01-01

    RA research reacto was not operated due to the prohibition issued in 1984 by the Government of Serbia. Three major tasks were finished in order to fulfill the licensing regulations about safety of nuclear facilities which is the condition for obtaining permanent operation licence. These projects involved construction of the emergency cooling system, reconstruction of the existing special ventilation system, and renewal of the system for electric power supply of the reactor systems. Renewal of the RA reactor instrumentation system was initiated. Design project was done by the Russian Atomenergoeksport, and is foreseen to be completed by the end of 1988. The RA reactor safety report was finished in 1987. This annual report includes 8 annexes concerning reactor operation, activities of services and financial issues, and three special annexes: report on testing the emergency cooling system, report on renewal of the RA reactor and design specifications for reactor renewal and reconstruction [sr

  16. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

  17. Twenty years of operation of Ljubljana's TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Dimic, V.

    1986-01-01

    Twenty years have now passed since the start of the TRIGA Mark II reactor in Ljubljana. The reactor was critical on May 31, 1966. The total energy produced until the end of May 1986 was 14.048 MWh or 585 MWd. For the first 14 years (until 1981) the yearly energy produced was about 600 MWh, since 1981 the yearly energy produced was 1000 MWh when a routine radioactive isotopes production started for medical use as well as other industrial applications, such as doping and irradiation with fast neutrons of silicon monocrystals, production of level indicators (irradiated cobalt wire), production of radioactive iridium for gamma-radiography, leak detection in pipes by sodium, etc. Besides these, applied research around the reactor is being conducted in the following main fields, where- many unique methods have been developed or have found their way into the local industry or hospitals: neutron radiography, neutron induced auto-radiography using solid state nuclear track detectors, nondestructive methods for assessment of nuclear burn-up, neutron dosimetry, calculation of core burn-up for the optimal in-core fuel management strategy. The solvent extraction method was developed for the everyday production of 99m Tc, which is the most widely used radionuclide in diagnostic nuclear medicine. The methods were developed for the production of the following isotopes: 18 F, 85m Kr, 24 Na, 82 Br, 64 Zn, 125 I. Neutron activation analysis represents one of the major usages for the TRIGA reactor. Basic research is being conducted in the following main fields: solid state physics (elastic and inelastic scattering of the neutrons), neutron dosimetry, neutron radiography, reactor physics and neutron activation analysis. The reactor is used very extensively as a main instrument in the Reactor Training Centre in Ljubljana where manpower training for our nuclear power plant and other organisations has been performed. Although the reactor was designed very carefully in order to be used for

  18. High-Performance Operating Systems

    DEFF Research Database (Denmark)

    Sharp, Robin

    1999-01-01

    Notes prepared for the DTU course 49421 "High Performance Operating Systems". The notes deal with quantitative and qualitative techniques for use in the design and evaluation of operating systems in computer systems for which performance is an important parameter, such as real-time applications......, communication systems and multimedia systems....

  19. Operational Safety Performance Indicators and Balanced Scorecard in HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Ahn, Guk-Hoon; Lee, Kye-Hong; Lim, In-Cheol; Kim, Hark-Rho

    2007-01-01

    Research reactors need an extensive basis for ensuring their safety. The importance of a safety management in nuclear facilities and activities has been emphasized. The safety activities in HANARO have been continuously conducted to enhance its safe operation. Last year, HANARO prepared two indicator sets to measure and assess the safety status of the reactor's operation and utilization. One is Safety Performance Indicators (SPI) and the other is Balanced Scorecard (BSC). Through reviewing these indicators, we can obtain the following information; - Plant safety status - Safety parameter trends - Safety information, for example, reactor operation status and radiation safety HANARO will continuously pursue the trends of SPI and BSC

  20. Artificial intelligence program in a computer application supporting reactor operations

    International Nuclear Information System (INIS)

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II

  1. Code on the safety of nuclear research reactors: Operation

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this publication is to provide the essential requirements and recommendations for the safe operation of research reactors, with emphasis on the supervisory and managerial aspects. However, the publication also provides some guidance and information on topics concerning all the organizations involved in operation. These objectives are expressed in terms of requirements and recommendations for the safe operation of research reactors. Emphasis is placed on the safety requirements that shall be met rather than on the ways in which they can be met. The requirements and recommendations may form the foundation necessary for a Member State to develop regulations and safety criteria for its research reactor programme.

  2. Modern control technology for improved nuclear reactor performance

    International Nuclear Information System (INIS)

    Oakes, L.C.

    1986-01-01

    One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level

  3. Density dependence of reactor performance with thermal confinement scalings

    International Nuclear Information System (INIS)

    Stotler, D.P.

    1992-03-01

    Energy confinement scalings for the thermal component of the plasma published thus far have a different dependence on plasma density and input power than do scalings for the total plasma energy. With such thermal scalings, reactor performance (measured by Q, the ratio of the fusion power to the sum of the ohmic and auxiliary input powers) worsens with increasing density. This dependence is the opposite of that found using scalings based on the total plasma energy, indicating that reactor operation concepts may need to be altered if this density dependence is confirmed in future research

  4. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  5. Development, operational experience and implications for future design of fast reactors in Western Europe

    International Nuclear Information System (INIS)

    Brandstetter, A.; Broomfield, A.M.; Saitcevsky, B.

    1990-01-01

    Over the past 30 years, the partners now collaborating in Europe on fast-reactor development have taken the technology from a theoretical possibility to an engineering reality. In that time there has been a progression from experimental zero-energy facilities followed by small power-producing engineering test reactors, to prototype reactors and a large commercial-size reactor. The paper describes the highlights of the reactor programmes in the partner countries and by example illustrates the experience gained from reactor operation and some of the principal activities in the supporting development programme. These include such topics as fuel performance, fast-neutron physics, liquid-metal thermal hydraulics, sodium chemistry, instrumentation and safety aspects. The paper concludes by summarizing some of the main objectives of the current development programme, which is directed to the support of the European Fast Reactor design being prepared by the European design and construction companies. (author)

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

    International Nuclear Information System (INIS)

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

    1976-07-01

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

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  8. Operating reactors licensing actions summary. Volume 5, No. 9

    International Nuclear Information System (INIS)

    1985-11-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  9. Operating reactors licensing actions summary. Volume 5, No. 8

    International Nuclear Information System (INIS)

    1985-10-01

    This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  10. Operating reactors licensing actions summary. Volume 4, No. 9

    International Nuclear Information System (INIS)

    1984-11-01

    This document is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the division of licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  11. Operating reactors licensing actions summary. Vol. 4, No. 2

    International Nuclear Information System (INIS)

    1984-04-01

    This summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Resource Management

  12. Development of operation management database for research reactors

    International Nuclear Information System (INIS)

    Zhang Xinjun; Chen Wei; Yang Jun

    2005-01-01

    An Operation Database for Pulsed Reactor has been developed on the platform for Microsoft visual C++ 6.0. This database includes four function modules, fuel elements management, incident management, experiment management and file management. It is essential for reactor security and information management. (authors)

  13. Reactor oscillator - Proposal of the organisation for oscillator operation; Reaktorski oscilator - Predlog organizacije rada na oscilatoru

    Energy Technology Data Exchange (ETDEWEB)

    Lolic, B; Loloc, B [Institute of Nuclear Sciences Boris Kidric, Laboratorija za fiziku reaktora, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

    The organizational structure for operating the reactor with the reactor oscillator describes the duties of the reactor operators; staff responsible for operating the oscillator who are responsible for measurements, preparation of the samples and further treatment of the obtained results.

  14. Physics experiments with the operating reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cullington, G R; King, D C

    1973-09-27

    Experimental techniques have been developed and used on Dragon to give consistent information on excess reactivity and shut down margin. The reactivity measurements have been correlated with the theoretical calculations and have led to improvements in the calculations. The methods used and the results obtained are accepted by the Safety Committee as sufficient evidence for compliance with the fuel loading safety rules. Although the reactor was not designed as an experimental facility, flux and dose measurements experiments have been successfully carried out. Mass flow and negative reactivity transient measurements have been carried out. These are valuable for demonstration of the flexibility of the reactor system and for giving confidence in theoretical calculations.

  15. Fuel deposits, chemistry and CANDU® reactor operation

    International Nuclear Information System (INIS)

    Roberts, J.G.

    2014-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU® reactor, the first being the Nuclear Power Demonstration - 2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channelled to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5. The difference being that during 'hot conditioning' of CANDU® heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  16. Ra reactor operation plan for 1976 - Annex 6

    International Nuclear Information System (INIS)

    Martinc, R.

    1976-01-01

    Contrary to previous years the RA reactor operation plan for 1976 is conditioned by more uncertainties than ever for the following reasons. It is planned for February to repair the components in order to eliminate the consequences of leaking from the vertical experimental channel VK-4 which happened in december 1975. Apart from this, measurements related to transfer of the core to highly enriched fuel under minimum power are foreseen to be performed in February. In April, it would be indispensable to locate the point of He gas leaking and its prevention. Core transfer to highly enriched fuel is planned for April for testing of the fuel elements under operating conditions and step by step increase of the power density up to the nominal power. Complete plan for core transfer would be available by the end of February 1976. It is planned to make-up for the mentioned delays from May-December 1976 [sr

  17. Simulation for Remote Operation for REX10 Nuclear Reactor

    International Nuclear Information System (INIS)

    Lee, Sim Won; Kim, Dong Su; Na, Man Gyun; Lee, Yoon Joon; Lee, Yeon Gun; Park, Goon Cherl

    2010-01-01

    The newly designed REX10 (Regional Energy Reactor, 10MWth) is an environmentally-friendly and stable small nuclear reactor for a small-scale reactor based Multi-purpose regional energy system. The REX10 has been developed to maintain system safety in order to be placed in densely populated region, island, etc. In addition, it is significantly hard to recruit many operation and maintenance personnel for small power reactors differently from usual commercial reactors because of its remote location and of economic reasons. In order to overcome these constraints, to decrease the operation cost by reducing operation and maintenance personnel, and to increase plant reliability through autonomous plant control, it is needed to design the control system of the small power reactors and to establish its unmanned remote operation system. In this study, the REX10 reactor core thermal power controller is designed by using a REX10 code analyzer. The remote control facility through man-machine interface (MMI) design and interface between programming languages was established and it was used to verify remote operation of REX10

  18. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1986

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1986-01-01

    In order to enable future reliable operation of the RA reactor, according to new licensing regulations, three major tasks started in 1984 were fulfilled: building of the new emergency system, reconstruction of the existing ventilation system, and reconstruction of the power supply system. Simultaneously in 1985/1986 renewal of the instrumentation and reconstruction of the system for handling and storage of the spent fuel in the reactor building have started. Design projects for these tasks are almost finished and the reconstruction of both systems is expected to be finished until 1988 and mid 1989 respectively. RA reactor Safety report was finished according to the recommendations of the IAEA. Investments in 1986 were used for 8000 kg of heavy water, maintenance of reactor systems and supply of new components, reconstruction of reactor systems. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  19. Reactor helium system, design specification, operation and handling

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-06-01

    Apart from detailed design specification of the helium cover gas system of the Ra reactor, this document includes description of the operating regime, instructions for manipulations in the system with the aim of achieving and maintaining stationary gas circulation [sr

  20. Operating history of U.S. nuclear power reactors

    International Nuclear Information System (INIS)

    1974-01-01

    The operating history of U. S. nuclear power plants through December 31, 1974 has been collected. Included are those nuclear reactor facilities which produce electricity, even if in token amounts, or which are part of a development program concerned with the generation of electricity through the use of a nuclear reactor as a heat source. The information is based on data furnished by facility operators. The charts are plotted in terms of cumulative thermal energy as a function of time. Since only those shutdowns of five days or more are shown, the charts do not give a detailed history of plant operation. They do, however, give an overview of the operating history of a variety of developmental and experimental nuclear power reactors. The data show the yearly gross generation of electricity for each U. S. nuclear plant and, for civilian power plants, information on reactor availability and plant capacity factor. (U.S.)

  1. RA reactor operation and maintenance; Pogon i odrzavanje reaktora RA

    Energy Technology Data Exchange (ETDEWEB)

    Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

    1963-02-15

    This volume includes the final report on RA reactor operation and utilization of the experimental facilities in 1962, detailed analysis of the system for heavy water distillation and calibration of the system for measuring the activity of the air.

  2. Safe operation of critical assemblies and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1960-09-15

    Some countries have accumulated considerable experience in the operation of these reactors and have in the process developed safe practices. On the other hand, other countries which have recently acquired, or will soon acquire, such reactors do not have sufficient background of experience with them to have developed full knowledge regarding their safe operation. In this situation, the International Atomic Energy Agency has considered that it would be useful to make available to all its Member States a set of recommendations on the safe operation of these reactors, based on the accumulated experience and best practices. The Director General accordingly nominated a Pane Ion Safe Operation of Critical Assemblies and Research Reactors to assist the Agency's Secretariat in drafting such recommendations

  3. Performance of an innovative multi-stage anaerobic reactor during ...

    African Journals Online (AJOL)

    Start-up of an anaerobic reactor is a relatively delicate process and depends on various factors such as wastewater composition, available inoculum, operating conditions and reactor configuration. Accordingly, systematized operational procedures are important, mainly during the start-up of an anaerobic reactor.

  4. Operation monitoring and protection method for nuclear reactor

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi.

    1995-01-01

    In an operation and monitoring method for a PWR-type reactor by using a tetra-sected neutron detector, axial off set is defined by neutron detector signals with respect to an average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core. A departure from nucleate boiling (DNBR) is represented by standardized signals, and the DNBR is calculated by using the axial off set of the average of the reactor core, the upper half of the reactor core, and the lower half of the reactor core, and they are graphically displayed. In addition, a thermal flow rate-water channel coefficient is also graphically displayed, and the DNBR and the thermal flow rate-water channel coefficient are restricted based on the display, to determine an allowable operation range. As a result, it is possible to provide an operation monitoring and protection method for nuclear reactor capable of reducing labors and frequencies for the change of protection system setting in a case of using a tetra-sected neutron detector disposed at the outside and, at the same time, protecting each of DNR and the highest linear power or the thermal water coefficient channel. (N.H.)

  5. A three-dimensional operational transient simulation of the CANDU core with typical reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae; Park, Kyung Seok; Park, Jong Woon [Institute for Advanced Engineering, Taejon (Korea, Republic of)

    1995-07-01

    This paper describes the results of simulation of a CANDU operational transient problem (re-startup after short shutdown) using the Coupled Reactor Kinetics(CRKIN) code developed previously with CANDU Reactor Regulating System (RRS) logic. The performance in the simulation is focused on investigating the behaviours of neutron power and regulating devices in accordance with the changes of xenon concentration following the operation of the RRS.

  6. Procedure for operating a heavy water cooled power reactor

    International Nuclear Information System (INIS)

    Rau, P.; Kumpf, H.

    1981-01-01

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

  7. EDF view on next generation reactor safety and operability issues

    International Nuclear Information System (INIS)

    Serviere, G.

    2002-01-01

    In the foreseeable future, EDF will have to compete in an economically de-regulated market. Nuclear currently accounts for more than 80% of the electricity generated by the company, and generation costs are quite competitive compared to that of other competing energies. It is so likely that nuclear units will remain the backbone of EDF generating fleet in the years to come. However, to remain a viable option for electricity generation in the longer term, nuclear will have to maintain both its cost-effectiveness and a very high safety level. This could seem quite straightforward considering the current situation where safety records are at an all time high and Operating and Maintenance costs are under tight control. In fact, it could be a real challenge. Competing fossil technologies progress and there is a concurrent trend to try and improve the performance of future nuclear units. However, in most cases, proposed designs depart from the well-known Light Water Reactor (LWR) technology. They are either new concepts or designs already tested in the past and modified to address some of their perceived drawbacks. Contrary to the prevailing situation where short-term alternatives like the EPR, the ABWR or the AP600 largely build upon experience gathered on operating units, most designs contemplated for implementation beyond 2020 or 2030 cannot be considered proven. Considering the above mentioned uncertainties, EDF have confirmed their preference for proven designs with higher outputs, such as the EPR. However, it would appear unreasonable to consider that new designs are doomed to fail: they could well turn out to be adequate for specific niches in a de-regulated market and provide reasonable alternatives for the utility. Nevertheless, for such an alternative to be considered, additional evidence is needed that utility preferences are reflected in the design, and that all potential technical issues have been identified, adequately addressed and resolved. Currently, EDF

  8. The IPR-R1 TRIGA Mark I Reactor in 39 years: Operations and general improvements

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Prado Fernandes, Marcio; Oliveira, Paulo Fernando; Alves de Amorim, Valter

    1999-01-01

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960. Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 kW, a new control rod mechanism, an aluminum tank for the reactor pool, an optimization in the pneumatic system, a new reactor control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. This paper describes the improvements made, the results obtained during the past 39 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe work. (author)

  9. Theoretical and experimental analysis of fast reactor fuel performance

    International Nuclear Information System (INIS)

    Kummerer, K.R.; Freund, D.; Steiner, H.

    1982-09-01

    In order to predict behavior, performance, and capability of prototypic fuel pins a standard operational scheme for the SNR-300 fast breeder reactor is established considering besides normal operation unscheduled power changes and shutdowns. The behavior during the whole lifetime is calculated using the updated SATURN codes and - for special conditions as power transients and skewed fuel rod power - the new TRANSIENT and TEXDIF codes. The results of these calculations are compared to experimental findings. It is demonstrated that the level of modeling and the knowledge of material properties under irradiation are sufficient for a quantitative description of the fuel pin performance under the above mentioned conditions. (orig.) [de

  10. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1984

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1984-12-01

    During the 1984 the reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981. Operation of the primary cooling system was changed in order to avoid appearance of the previously noticed aluminium oxyhydrate on the surface of the fuel element claddings. The new cooling regime enabled more efficient heavy water purification. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks are planned: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. Financing of the planned activities will be partly covered by the IAEA. this Part I of the report includes 8 Annexes describing in detail the reactor operation, and 6 special papers dealing with the problems of reactor operation and utilization

  11. Reactor operator: Training for the job while earning college credit

    International Nuclear Information System (INIS)

    Murdick, S.A.

    1988-01-01

    The nuclear industry is looking for ways to maximize the dollars spent to train licensed reactor operators and other personnel and, at the same time, upgrade their educational level. The prospects of college credit and/or degree requirements imposed by the US Nuclear Regulatory Commission have provided a significant driving force behind this search. The task is complicated, however, because shift schedules do not permit reactor operators to pursue higher education through the traditional classroom route, and the need for plant-specific training and requalification programs dictate against uniformly adapting college-based courses for training use. The National Program on Noncollegiate Sponsored Instruction (National PONSI) has been of considerable help to the nuclear industry in meeting these challenges. Through its college credit recommendation service, National PONSI has assessed the comparability of certain industry training activities to college-level instruction and has been instrumental in gaining academic recognition of these activities. The program has become a vital means for the industry to achieve its dual mission of preparing employees to successfully perform their jobs and providing them with ways to obtain college degrees in the shortest possible time

  12. Reactor operator: Training for the job while earning college credit

    Energy Technology Data Exchange (ETDEWEB)

    Murdick, S.A.

    1988-01-01

    The nuclear industry is looking for ways to maximize the dollars spent to train licensed reactor operators and other personnel and, at the same time, upgrade their educational level. The prospects of college credit and/or degree requirements imposed by the US Nuclear Regulatory Commission have provided a significant driving force behind this search. The task is complicated, however, because shift schedules do not permit reactor operators to pursue higher education through the traditional classroom route, and the need for plant-specific training and requalification programs dictate against uniformly adapting college-based courses for training use. The National Program on Noncollegiate Sponsored Instruction (National PONSI) has been of considerable help to the nuclear industry in meeting these challenges. Through its college credit recommendation service, National PONSI has assessed the comparability of certain industry training activities to college-level instruction and has been instrumental in gaining academic recognition of these activities. The program has become a vital means for the industry to achieve its dual mission of preparing employees to successfully perform their jobs and providing them with ways to obtain college degrees in the shortest possible time.

  13. In-situ inspection of grooves in reactor tube sheet using a remotely operated cast impression taking device

    International Nuclear Information System (INIS)

    Rajendran, S.; Ramakumar, M.S.

    1996-01-01

    Utmost importance is given to the in-service inspection of critical components of a reactor to ensure its reliable performance during the reactor operation. This paper describes a cast taking device using cold setting resin to take impression of the grooves being made in the tube sheet for sparger tube installation in pressurised heavy water reactor. (author)

  14. Dry cooling tower operating experience in the LOFT reactor

    International Nuclear Information System (INIS)

    Hunter, J.A.

    1980-01-01

    A dry cooling tower has been uniquely utilized to dissipate heat generated in a small experimental pressurized water nuclear reactor. Operational experience revealed that dry cooling towers can be intermittently operated with minimal wind susceptibility and water hammer occurrences by cooling potential steam sources after a reactor scram, by isolating idle tubes from the external atmosphere, and by operating at relatively high pressures. Operating experience has also revealed that tube freezing can be minimized by incorporating the proper heating and heat loss prevention features

  15. Fuel deposits, chemistry and CANDU reactor operation

    International Nuclear Information System (INIS)

    Roberts, J.G.

    2013-01-01

    'Hot conditioning' is a process which occurs as part of commissioning and initial start-up of each CANDU reactor, the first being the Nuclear Power Demonstration-2 reactor (NPD). Later, understanding of the cause of the failure of the Pickering Unit 1 G16 fuel channel led to a revised approach to 'hot conditioning', initially demonstrated on Bruce Unit 5, and subsequently utilized for each CANDU unit since. The difference being that during 'hot conditioning' of CANDU heat transport systems fuel was not in-core until Bruce Unit 5. The 'hot conditioning' processes will be briefly described along with the consequences to fuel. (author)

  16. Operation control equipment for BWR type reactor

    International Nuclear Information System (INIS)

    Izumi, Masayuki; Takeda, Renzo.

    1981-01-01

    Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

  17. Method of operating a nuclear reactor

    International Nuclear Information System (INIS)

    Gyorey, G.L.; Parkos, G.R.; Roupe, G.A.; Thomson, O.A.; Crowther, R.L.

    1979-01-01

    The invention concerns the configuration of control rods in the lattice of the reactor core, as well as an instruction on the sequence of with drawal for the control rods, arranged in groups, in order to achieve for the control rod reactivity of the control rods remaining in the reactor core to adopt the lowest possible value. The rods are combined in several 3 x 3 matrices which in their turn are grouped into two networks. The groups are moved successively according to a specified schedule. There can be achieved maximum control rod reactivities between 0.025 and 0.035 (referred to the totally withdrawn state). (RW) 891 RW/RW 892 MKO [de

  18. Operating experience feedback from safety significant events at research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shokr, A.M. [Atomic Energy Authority, Abouzabal (Egypt). Egypt Second Research Reactor; Rao, D. [Bhabha Atomic Research Centre, Mumbai (India)

    2015-05-15

    Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

  19. L-Reactor operation, Savannah River Plant: environmental assessment

    International Nuclear Information System (INIS)

    1982-08-01

    The purpose of this document is to assess the significance of the effects on the human environment of the proposed resumption of L-reactor operation at the Savannah River Plant, scheduled for October 1983. The discussion is presented under the following section headings: need for resumption of L-Reactor operations and purpose of this environmental assessment; proposed action and alternative; affected environment (including, site location and description, land use, historic and archeological resources, socioeconomic and community characteristics, geology and seismology, hydrology, meteorology and climatology, ecology, and radiation environment); environmental consequences; summary of projected L-Reactor releases and impacts; and Federal and State permits and approval. The three appendices are entitled: radiation dose calculation methods and assumptions; floodplain/wetlands assessment - L-Reactor operations; and, conversion table. A list of references is included at the end of each chapter

  20. Operating experience feedback from safety significant events at research reactors

    International Nuclear Information System (INIS)

    Shokr, A.M.

    2015-01-01

    Operating experience feedback is an effective mechanism to provide lessons learned from the events and the associated corrective actions to prevent recurrence of events, resulting in improving safety in the nuclear installations. This paper analyzes the events of safety significance that have been occurred at research reactors and discusses the root causes and lessons learned from these events. Insights from literature on events at research reactors and feedback from events at nuclear power plants that are relevant to research reactors are also presented along with discussions. The results of the analysis showed the importance of communication of safety information and exchange of operating experience are vital to prevent reoccurrences of events. The analysis showed also the need for continued attention to human factors and training of operating personnel, and the need for establishing systematic ageing management programmes of reactor facilities, and programmes for safety management of handling of nuclear fuel, core components, and experimental devices.

  1. Completely automated nuclear reactors for long-term operation

    International Nuclear Information System (INIS)

    Teller, E.; Ishikawa, M.; Wood, L.

    1996-01-01

    The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant's operational and post-operational life. These reactors are proposed to be situated in suitable environments at ∼100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant's heat engine and electrical generator subsystems are located above-ground

  2. Operation characteristics and conditions of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

    1994-01-01

    The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

  3. Research nuclear reactor RA - Annual Report 1975. Operation and maintenance

    International Nuclear Information System (INIS)

    Martinc, R.

    1976-01-01

    The plan for 1975 was successfully fulfilled. This is reflected in research related to improvement of operating properties of the RA reactor, mostly due to the effort of the RA staff employed in operation and maintenance of the reactor. Fuel saving achieved by this activity amounted to about 38% (80% enriched fuel). Preliminary work is done, concerned with new reactor core with highly enriched fuel. This is a significant saving as well. New fuel elements have arrived at the end of this year. It is going to enable increase of neutron flux by 50% without changing the nominal operating power. The possibility of further improvement of the reactor are analyzed, to enable material testing and production of radioactive sources. Mid term plan for reactor operation was made according to this analysis. It is planned to further increase the neutron flux in isolated smaller zones, and building new experimental loops with cooling and fast neutron converters. Much was done to increase the safety level of reactor operation and preparing the safety report [sr

  4. Development of Abnormal Operating Strategies for Station Blackout in Shutdown Operating Mode in Pressurized Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Duk-Joo; Lee, Seung-Chan; Sung, Je-Joong; Ha, Sang-Jun [KHNP CRI, Daejeon (Korea, Republic of); Hwang, Su-Hyun [FNC Tech. Co., Yongin (Korea, Republic of)

    2016-10-15

    Loss of all AC power is classified as one of multiple failure accident by regulatory guide of Korean accident management program. Therefore we need develop strategies for the abnormal operating procedure both of power operating and shutdown mode. This paper developed abnormal operating guideline for loss of all AC power by analysis of accident scenario in pressurized water reactor. This paper analyzed the loss of ultimate heat sink (LOUHS) in shutdown operating mode and developed the operating strategy of the abnormal procedure. Also we performed the analysis of limiting scenarios that operator actions are not taken in shutdown LOUHS. Therefore, we verified the plant behavior and decided operator action to taken in time in order to protect the fuel of core with safety. From the analysis results of LOUHS, the fuel of core maintained without core uncovery for 73 minutes respectively for opened RCS states after the SBO occurred. Therefore, operator action for the emergency are required to take in 73 minutes for opened RCS state. Strategy is to cooldown by using spent fuel pool cooling system. This method required to change the plant design in some plant. In RCS boundary closed state, first abnormal operating strategy in shutdown LOUHS is first abnormal operating strategy in shutdown LOUHS is to remove the residual heat of core by steam dump flow and auxiliary feedwater of SG.

  5. RA research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1985

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1985-01-01

    According to the plan, RA reactor was to be in operation in mid September 1985. But, since the building of the emergency cooling system, nor the reconstruction of the existing special ventilation system were not finished until the end of August reactor was not operated during 1985. During the previous four years reactor operation was limited by the temporary operating license issued by the Committee of Serbian ministry for health and social care, which was cancelled in August 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. This temporary license has limited the reactor power to 2 MW from 1981-1984. Control and maintenance of the reactor instrumentation and tools was done regularly but dependent on the availability of the spare parts. In order to enable future reliable operation of the RA reactor, according to new licensing regulations, during 1984, three major tasks have started: building of the new emergency system, reconstruction of the existing ventilation system, and renewal of the reactor instrumentation. IAEA has approved the amount of 1,300,000 US dollars for the renewal of the instrumentation [sr

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1968-04-15

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

  7. The evaluation of operator reliability factors on power reactor

    International Nuclear Information System (INIS)

    Karlina, Itjeu; Supriatna, Piping; W, Suharyo; Santosa, Kussigit; Darlis; S, Bambang; Y, Sasongko

    1999-01-01

    The sophisticated technology system was not assured the reliability system itself because it has contained a part of human dependence affected successfully of reactor operation either how work smoothly and safe or failure ac cured and then accident appears promptly. The evaluation of operator reliability factor on ABWR power reactor has been carried out which consist of criterion skill and workload according to NUREG/CR-2254, NUREG/CR-4016 and NUREG-0835 the reactor operation reliability emphasize to the operator are synergic between skill and workload themselves. The employee's skill will affect to the type and level of their tasks. The operator's skill depend on education and experiences, position or responsibility of tasks, physical conditions (age uninvalid of physic/mental

  8. Operation and maintenance of the RA reactor, Report on operation in 1966, task 6.08

    International Nuclear Information System (INIS)

    Milosevic, D. and others

    1966-12-01

    During 1966, RA reactor was operated at nominal power for 196 days and 28 days at lower power levels. The total production was 31015 MWh which is 3% higher than planned. Practically there were no discrepancies from the plan. The reactor was used for irradiation and experiments according to the demand of 644 users, of which 516 were from the Institute. This report contains detailed data about reactor power and experiments performed in 1966, and a chart showing discrepancies from the plan. data concerned about the utilization of reactor experimental channels are included as well. Total number of safety shutdowns was 37, of which 20 were caused by power cuts, 10 due to voltage instabilities, 4 due to human error and 3 due to failures of instrumentation. These data are compared to safety shutdown data of previous few years. There have been no accidents which could cause any significant effects. The incidents occurred during this year are described in detail. Detailed data concerning exposure of the personnel are part of this report

  9. Self-operation type power control device for nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kanbe, Mitsuru.

    1993-07-23

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.).

  10. Self-operation type power control device for nuclear reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru.

    1993-01-01

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

  11. Extending the maximum operation time of the MNSR reactor.

    Science.gov (United States)

    Dawahra, S; Khattab, K; Saba, G

    2016-09-01

    An effective modification to extend the maximum operation time of the Miniature Neutron Source Reactor (MNSR) to enhance the utilization of the reactor has been tested using the MCNP4C code. This modification consisted of inserting manually in each of the reactor inner irradiation tube a chain of three polyethylene-connected containers filled of water. The total height of the chain was 11.5cm. The replacement of the actual cadmium absorber with B(10) absorber was needed as well. The rest of the core structure materials and dimensions remained unchanged. A 3-D neutronic model with the new modifications was developed to compare the neutronic parameters of the old and modified cores. The results of the old and modified core excess reactivities (ρex) were: 3.954, 6.241 mk respectively. The maximum reactor operation times were: 428, 1025min and the safety reactivity factors were: 1.654 and 1.595 respectively. Therefore, a 139% increase in the maximum reactor operation time was noticed for the modified core. This increase enhanced the utilization of the MNSR reactor to conduct a long time irradiation of the unknown samples using the NAA technique and increase the amount of radioisotope production in the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  12. The human factor in nuclear reactor operation

    International Nuclear Information System (INIS)

    Bertron, L.

    1982-05-01

    The principal operating characteristics of nuclear power plants are summarized. A description of major hazards relating to operator fallibility in normal and abnormal operating conditions is followed by a specific analysis of control room hazards, shift organization and selection and training of management personnel

  13. Designing for a safe response to operational and severe accident initiators in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1994-01-01

    A method is described for optimizing the plant control strategy for a liquid metal reactor with respect to safety margins sustained in unprotected upset events. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  14. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    Fukutomi, S.; Takigawa, Y.

    1992-01-01

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  15. Performance demonstration experience for reactor pressure vessel shell ultrasonic testing

    International Nuclear Information System (INIS)

    Zado, V.

    1998-01-01

    The most ultrasonic testing techniques used by many vendors for pressurized water reactor (PWR) examinations were based on American Society of Mechanical Engineers 'Boiler and Pressurized Vessel Code' (ASME B and PV Code) Sections XI and V. The Addenda of ASME B and PV Code Section XI, Edition 1989 introduced Appendix VIII - 'Performance Demonstration for Ultrasonic Examination Systems'. In an effort to increase confidence in performance of ultrasonic testing of the operating nuclear power plants in United States, the ultrasonic testing performance demonstration examination of reactor vessel welds is performed in accordance with Performance Demonstration Initiative (PDI) program which is based on ASME Code Section XI, Appendix VIII requirements. This article provides information regarding extensive qualification preparation works performed prior EPRI guided performance demonstration exam of reactor vessel shell welds accomplished in January 1997 for the scope of Appendix VIII, Supplements IV and VI. Additionally, an overview of the procedures based on requirements of ASME Code Section XI and V in comparison to procedure prepared for Appendix VIII examination is given and discussed. The samples of ultrasonic signals obtained from artificial flaws implanted in vessel material are presented and results of ultrasonic testing are compared to actual flaw sizes. (author)

  16. RA reactor operation and maintenance in 1989, Part 1

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Zivotic, Z.; Majstorovic, D.; Sanovic, V.

    1989-01-01

    During the previous period RA reactor was not operated because the Committee of Serbian ministry for health and social care has cancelled the operation licence in July 1984. The reason was the non existing emergency cooling system and lack of appropriate filters in the special ventilation system. The following major tasks were fulfilled: building of the new emergency cooling system, reconstruction of the existing ventilation system, and renewal of the power supply system. Project concerned with renewal of RA reactor complete instrumentation was started at the end of 1988. Contract was signed between the IAEA and Soviet Atomenergoexport for supplying the new instrumentation for the RA reactor. Project concerned with increase of the storage space and the efficiency of handling the spent fuel elements has started in 1988. In 1989, device for water purification designed by the reactor staff started operation and spent fuel handling equipment is being mounted. Training of the existing personnel and was done regularly, but the new staff has no practical training since the reactor is not operated. Lack of financial support influenced strongly the status of RA reactor [sr

  17. RA Research reactor, Part 1, Operation and maintenance of the RA nuclear reactor for 1988

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Cupac, S.; Sulem, B.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1988-01-01

    According to the action plan for 1988, operation of the RA reactor should have been restarted in October, but the operating license was not obtained. Control and maintenance of the reactor components was done regularly and efficiently dependent on the availability of the spare parts. The major difficulty was maintenance of the reactor instrumentation. Period of the reactor shutdown was used for repair of the heavy water pumps in the primary coolant loop. With the aim to ensure future safe and reliable reactor operation, action were started concerning renewal of the reactor instrumentation. Design project was done by the soviet company Atomenergoeksport. The contract for constructing this equipment was signed, and it is planned that the equipment will be delivered by the end of 1990. In order to increase the space for storage of the irradiated fuel elements and its more efficient usage, projects were started concerned with reconstruction of the existing fuel handling equipment, increase of the storage space and purification of the water in the fuel storage pools. These projects are scheduled to be finished in mid 1989. This report includes 8 annexes concerning reactor operation, activities of services and financial issues [sr

  18. Comparative Study on Research Reactor Operation Requirements of IAEA and Korea

    International Nuclear Information System (INIS)

    Yune, Young Gill; Chang, Woo Joon; Lee, Jinho

    2014-01-01

    The IAEA safety requirements represent a consensus view of the IAEA's member states and encompass the recommendations of international experts for the safety of the research reactor. The IAEA safety standards are acknowledged as a global standard by most countries in the world. Therefore, it is desirable to harmonize domestic safety requirements with those of the IAEA. Also, since the IAEA safety requirements include international endeavors to strengthen nuclear safety and to apply the latest nuclear safety technologies to nuclear facilities, a comparative analysis of the safety requirements of the IAEA and Korea for the research reactor will be beneficial to obtaining the insights to improve regulations of the research reactor in Korea. For the reason, a comparative study has been conducted for the research reactor operation safety requirements of the IAEA and Korea in this paper. This paper briefly introduced operation safety requirements in the NS-R-4, analysis methods of this study, and the analysis results. A comparative study has been performed on research reactor operation safety requirements of the IAEA and Korea. This study has identified some gaps in operation safety requirements for the NPP of Korea, comparing with those for the research reactor of the IAEA in the following areas: periodic safety assessment, utilization and modification of the reactor, extended shutdown, inspection/testing/maintenance, operating organization, emergency planning, and commissioning

  19. Plant operator performance evaluation system

    International Nuclear Information System (INIS)

    Ujita, Hiroshi; Fukuda, Mitsuko; Kubota, Ryuji.

    1989-01-01

    A plant operator performance evaluation system to analyze plant operation records during accident training and to identify and classify operator errors has been developed for the purpose of supporting realization of a training and education system for plant operators. A knowledge engineering technique was applied to evaluation of operator behavior by both even-based and symptom-based procedures, in various situations including event transition due to multiple failures or operational errors. The system classifies the identified errors as to their single and double types based on Swain's error classification and the error levels reflecting Rasmussen's cognitive level, and it also evaluates the effect of errors on plant state and then classifies error influence, using 'knowledge for phenomena and operations', as represented by frames. It has additional functions for analysis of error statistics and knowledge acquisition support of 'knowledge for operations'. The system was applied to a training analysis for a scram event in a BWR plant, and its error analysis function was confirmed to be effective by operational experts. (author)

  20. Design of a periodically operated SCR reactor

    International Nuclear Information System (INIS)

    Kotter, M.; Lintz, H.G.; Turek, T.

    1993-01-01

    A new NO x abatement process uses the rotating Ljungstroem air heater of the power plant for the selective catalytic reduction (SCR) of nitrogen monoxide with ammonia. For this purpose the air heater elements are covered by a catalytically active layer. The transformation can be carried out by simple replacement of the original air heater elements. Thus nitrogen monoxide control is possible without requiring major modifications of existing power plant equipment. Two oxidic catalysts have been developed to be employed in the different temperature sections of the air heater. The activity of the catalysts has been quantified with the aid of laboratory scale experiments. The results can be described using a simple expression for the rate of the chemical reaction. NO conversion and NH 3 slip to be expected in a catalytically active Ljungstroem heat exchanger are calculated with a reactor model taking into account the gas phase mass transfer resistances. The calculations show that the proposed device can be used if the NO concentration in the flue gas does not exceed 300 ppm. Recently Kraftanlagen AG, Heidelberg, installed a catalyst air heater system at Mandalay Generating Station in Oxnard, California. The comparison of the predicted results with preliminary experimental data proves the validity of the chosen reactor model. Under the given conditions NO conversions of more than 60% can be achieved maintaining the NH 3 slip below the specified value of 10 ppm. (orig.). 19 figs., 35 refs [de

  1. Performance Monitoring for Nuclear Safety Related Instrumentation at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2015-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on performance monitoring for nuclear safety related instrumentation in TRIGA PUSPATI Reactor (RTP) of based on various parameter of reactor safety instrument channel such as log power, linear power, Fuel temperature, coolant temperature will take into consideration. Methodology of performance on estimation and monitoring is to evaluate and analysis of reactor parameters which is important of reactor safety and control. And also to estimate power measurement, differential of log and linear power and fuel temperature during reactor start-up, operation and shutdown .This study also focus on neutron power fluctuation from fission chamber during reactor start-up and operation. This work will present result of performance monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that performance of nuclear safety related instrumentation will improved the reactor control and safety parameter during reactor start-up, operation and shutdown. (author)

  2. Fluidized-bed reactors processes and operating conditions

    CERN Document Server

    Yates, John G

    2016-01-01

    The fluidized-bed reactor is the centerpiece of industrial fluidization processes. This book focuses on the design and operation of fluidized beds in many different industrial processes, emphasizing the rationale for choosing fluidized beds for each particular process. The book starts with a brief history of fluidization from its inception in the 1940’s. The authors present both the fluid dynamics of gas-solid fluidized beds and the extensive experimental studies of operating systems and they set them in the context of operating processes that use fluid-bed reactors. Chemical engineering students and postdocs as well as practicing engineers will find great interest in this book.

  3. Cognitive Performance in Operational Environments

    Science.gov (United States)

    Russo, Michael; McGhee, James; Friedler, Edna; Thomas, Maria

    2005-01-01

    Optimal cognition during complex and sustained operations is a critical component for success in current and future military operations. "Cognitive Performance, Judgment, and Decision-making" (CPJD) is a newly organized U.S. Army Medical Research and Materiel Command research program focused on sustaining operational effectiveness of Future Force Warriors by developing paradigms through which militarily-relevant, higher-order cognitive performance, judgment, and decision-making can be assessed and sustained in individuals, small teams, and leaders of network-centric fighting units. CPJD evaluates the impact of stressors intrinsic to military operational environments (e.g., sleep deprivation, workload, fatigue, temperature extremes, altitude, environmental/physiological disruption) on military performance, evaluates noninvasive automated methods for monitoring and predicting cognitive performance, and investigates pharmaceutical strategies (e.g., stimulant countermeasures, hypnotics) to mitigate performance decrements. This manuscript describes the CPJD program, discusses the metrics utilized to relate militarily applied research findings to academic research, and discusses how the simulated combat capabilities of a synthetic battle laboratory may facilitate future cognitive performance research.

  4. Simulator platform for fast reactor operation and safety technology demonstration

    International Nuclear Information System (INIS)

    Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-01-01

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  5. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  6. Optimal processor for malfunction detection in operating nuclear reactor

    International Nuclear Information System (INIS)

    Ciftcioglu, O.

    1990-01-01

    An optimal processor for diagnosing operational transients in a nuclear reactor is described. Basic design of the processor involves real-time processing of noise signal obtained from a particular in core sensor and the optimality is based on minimum alarm failure in contrast to minimum false alarm criterion from the safe and reliable plant operation viewpoint

  7. Some particular problems put by operating experimental reactors

    International Nuclear Information System (INIS)

    Candiotti, C.; Mabeix, R.; Uguen, R.

    1960-01-01

    On basis of a six years experience in operating research reactors, the authors explain, first, the difference in their utilization between these piles and another similar ones and, after, in consequence, they set off corresponding servitudes. These servitudes put very particular problems in operating itself, maintenance, modifications or additions on these apparatus. (author) [fr

  8. Decentralization of operating reactor licensing reviews: NRR Pilot Program

    International Nuclear Information System (INIS)

    Hannon, J.N.

    1984-07-01

    This report, which has incorporated comments received from the Commission and ACRS, describes the program for decentralization of selected operating reactor licensing technical review activities. The 2-year pilot program will be reviewed to verify that safety is enhanced as anticipated by the incorporation of prescribed management techniques and application of resources. If the program fails to operate as designed, it will be terminated

  9. Performing a local barrier operation

    Science.gov (United States)

    Archer, Charles J; Blocksome, Michael A; Ratterman, Joseph D; Smith, Brian E

    2014-03-04

    Performing a local barrier operation with parallel tasks executing on a compute node including, for each task: retrieving a present value of a counter; calculating, in dependence upon the present value of the counter and a total number of tasks performing the local barrier operation, a base value, the base value representing the counter's value prior to any task joining the local barrier; calculating, in dependence upon the base value and the total number of tasks performing the local barrier operation, a target value of the counter, the target value representing the counter's value when all tasks have joined the local barrier; joining the local barrier, including atomically incrementing the value of the counter; and repetitively, until the present value of the counter is no less than the target value of the counter: retrieving the present value of the counter and determining whether the present value equals the target value.

  10. Training reactor operators and shift supervisors

    International Nuclear Information System (INIS)

    Schwarz, O.

    1980-01-01

    To establish a central institution run by power plant operators to harmonize the training of power plant operating personnel was raised, and put into practice, quite early in the Federal Republic of Germany. A committee devoted to training plant crews, which had been set up by the organizations of German electricity utilities responsible for operating power plants, was changed into a Kraftwerksschule e.V. (Power Plant School) in 1963. This school runs training courses, along standard lines, for operating personnel of thermal power plants, especially for operators and power plant supervisors, in close cooperation with power plant operators. As the peaceful utilization of nuclear energy expanded, also the training of nuclear power plant operators was included in 1969. Since September 1977, the center has had a simulator of a PWR nuclear power plant, since January 1978 also that of a BWR plant available for training purposes. Besides routine operation the trainees also learn to control those incidents which occur only very rarely in real nuclear power plants. (orig./UA) [de

  11. Reactivity monitoring during reactor-reloading operations

    International Nuclear Information System (INIS)

    Baumann, N.P.; Ahlfeld, C.F.; Ridgely, G.C.

    1983-01-01

    At the Savannah River Plant (SRP) reloading operations during shutdown present special considerations in reactivity monitoring and control. Large reactivity changes may occur during reloading operations because of the heterogeneous nature of some core designs. This paper describes an improved monitoring system

  12. The use of computer graphics to assist in-reactor operations

    International Nuclear Information System (INIS)

    Smith, A.L.

    1995-01-01

    A number of complex inspection and repair tasks have been carried out inside Nuclear Electric's gas cooled reactors using multi-link manipulators. The paper describes how computer graphics are integrated into the manipulator system and perform a key role during many phases of a project. Graphics simulation is used to assist with conceptual design, detailed path planning, rehearsals, quality assurance, demonstrations and, critically, as an on-line tool to assist the real-time operation of the equipment whilst performing complex in-reactor tasks. The techniques used are illustrated by recent examples of use to assist the inspection and repair of in-reactor components. (Author)

  13. Operation of the SLOWPOKE-2 reactor in Jamaica

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C.N.; Lalor, G.C.; Vuchkov, M.K. [University of the West Indies, Kingston (Jamaica)

    2001-07-01

    Over the past sixteen years lCENS has operated a SLOWPOKE 2 nuclear reactor almost exclusively for the purpose of neutron activation analysis. During this period we have adopted a strategy of minimum irradiation times while optimizing our output in an effort to increase the lifetime of the reactor core and to maintaining fuel integrity. An inter-comparison study with results obtained with a much larger reactor at IPEN has validated this approach. The parameters routinely monitored at ICENS are also discussed and the method used to predict the next shim adjustment. (author)

  14. Method of operating a water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Lysell, G.

    1975-01-01

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

  15. Study on operational aspect of natural circulation HLMC reactor (1)

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Cahalan, J.E.; Spencer, B.W.

    2000-08-01

    The concept of a heavy liquid metal cooled fast reactor that achieves 100% natural circulation heat removal from the core has the potential to attain improved cost competitiveness through extreme simplification, proliferation resistance, and heightened passive safety. The concept offers the potential for simplifications in plant control strategies wherein inherent reactor feedbacks may restore balance between energy release and heat removal from the reactor during operation as well as providing passive reactivity shutdown in the event of transients involving failure to scram. This study was initiated to evaluate the operational characteristics of the 100% natural circulation reactor under normal and transient states using a plant dynamics analysis computer code and to seek design and operational optimization of the concept. In the current Phase I of the project, the stage for the overall study has been prepared. A coupled thermal hydraulics-kinetics plant dynamics analysis code has been developed/modified that has the capabilities to calculate operational and accident transients. Code input has been prepared for the heavy liquid metal cooled natural circulation reactor concept. A preliminary analysis using the plant dynamics code and its input to calculate three illustrative cases relevant to initial startup, shutdown following long-term operation, and change in turbine load demonstrates the capability to analyze typical transient cases. (author)

  16. Operation experience with the TRIGA reactor Wien 2004

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2004-01-01

    The TRIGA Mark-II reactor in Vienna is now in operation for more than 42 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent, the number of students participating in practical exercises has strongly increased, and also training courses for outside groups such as the IAEA or for the 2004 Eugene Wigner Course are using the reactor, because it is the only TRIGA reactor remaining in Austria. Therefore, there is no need for decommissioning and it is intended to operate it as long as possible into the next decade. Nevertheless, in early 2004 it was decided to prepare a report on a decommissioning procedure for a typical TRIGA Mark II reactor which lists the volumes, the activity and the weight of individual materials such as concrete, aluminium, stainless steel, graphite and others which will accumulate during this process (a summary of possible activated and contaminated materials and the activity of a single TRIGA fuel element as a function of fuel type and decay time in Bq is presented). The status of the reactor (instrumentation, fuel elements, cooling circuit, ventilation system, re-inspection and maintenance program, cost/benefit) is outlined. (nevyjel)

  17. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  18. PERFORMANCE IMPROVEMENT OF A CHEMICAL REACTOR BY NONLINEAR NATURAL OSCILLATIONS

    NARCIS (Netherlands)

    RAY, AK

    1995-01-01

    The dynamic behaviour of two coupled continuous stirred tank reactors in sequence is studied when the first reactor is being operated under limit cycle regimes producing self-sustained natural oscillations. The periodic output from the first reactor is then used as a forced input into the second

  19. Operation of staged membrane oxidation reactor systems

    Science.gov (United States)

    Repasky, John Michael

    2012-10-16

    A method of operating a multi-stage ion transport membrane oxidation system. The method comprises providing a multi-stage ion transport membrane oxidation system with at least a first membrane oxidation stage and a second membrane oxidation stage, operating the ion transport membrane oxidation system at operating conditions including a characteristic temperature of the first membrane oxidation stage and a characteristic temperature of the second membrane oxidation stage; and controlling the production capacity and/or the product quality by changing the characteristic temperature of the first membrane oxidation stage and/or changing the characteristic temperature of the second membrane oxidation stage.

  20. DUPIC fuel performance from reactor physics viewpoint

    International Nuclear Information System (INIS)

    Choi, H.; Rhee, B.W.; Park, H.

    1995-01-01

    A preliminary study was performed for the evaluation of Stress Corrosion Cracking (SCC) parameters of nominal DUPIC fuel in CANDU reactor. For the reference 2-bundle shift refueling scheme, the predicted ramped power and power increase of the 43-element DUPIC fuel in the equilibrium core are below the SCC thresholds of CANDU natural uranium fuel. For 4-bundle shift refueling scheme, the envelope of element ramped power and power increase upon refueling are 8% and 44% higher than those of 2-bundle shift refueling scheme on the average, respectively, and both schemes are not expected to cause SCC failures. (author)

  1. Modified-open fuel cycle performance with breed-and-burn advanced reactor concepts

    International Nuclear Information System (INIS)

    Heidet, Florent; Kim, Taek K.; Taiwo, Temitope A.

    2011-01-01

    Recent advances in fast reactor designs enable significant increase in the uranium utilization in an advanced fuel cycle. The category of fast reactors, collectively termed breed-and-burn reactor concepts, can use a large amount of depleted uranium as fuel without requiring enrichment with the exception of the initial core critical loading. Among those advanced concepts, some are foreseen to operate within a once-through fuel cycle such as the Traveling Wave Reactor, CANDLE reactor or Ultra-Long Life Fast Reactor, while others are intended to operate within a modified-open fuel cycle, such as the Breed-and-Burn reactor and the Energy Multiplier Module. This study assesses and compares the performance of the latter category of breed-and-burn reactors at equilibrium state. It is found that the two reactor concepts operating within a modified-open fuel cycle can significantly improve the sustainability and security of the nuclear fuel cycle by decreasing the uranium resources and enrichment requirements even further than the breed-and-burn core concepts operating within the once-through fuel cycle. Their waste characteristics per unit of energy are also found to be favorable, compared to that of currently operating PWRs. However, a number of feasibility issues need to be addressed in order to enable deployment of these breed-and-burn reactor concepts. (author)

  2. Performance tests for integral reactor nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Dong-Seong; Yim, Jeong-Sik; Lee, Chong-Tak; Kim, Han-Soo; Koo, Yang-Hyun; Lee, Byung-Ho; Cheon, Jin-Sik; Oh, Je-Yong

    2006-02-15

    An integral type reactor SMART plans to utilize metallic Zr-U fuel which is Zr-based alloy with 34{approx}38 wt% U. In order to verify the technologies for the design and manufacturing of the fuel and get a license, performance tests were carried out. Experimental Fuel Assembly (EFA) manufactured in KAERI is being successfully irradiated in the MIR reactor of RIAR from September 4 2004, and it has achieved burnup of 0.21 g/cc as of January 25 2006. Thermal properties of irradiated Zr-U fuel were measured. Up to the phase transformation temperature, thermal diffusivity increased linearly in proportion to temperature. However its dependence on the burnup was not significant. RIA tests with 4 unirradiated Zr-U fuel rods were performed in Kurchatov Institute to establish a safety criterion. In the case of the un-irradiated Zr-U fuel, the energy deposition during the control rod ejection accident should be less than 172 cal/g to prevent the failure accompanying fuel fragmentation and dispersal. Finally the irradiation tests of fuel rods have been performed at HANARO. The HITE-2 test was successfully completed up to a burnup of 0.31 g/cc. The HITE-3 test began in February 2004 and will be continued up to a target burnup of 0.6 g/cc.

  3. Fuel performance and fission product behaviour in gas cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport. Refs, figs, tabs.

  4. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    1997-11-01

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  5. Artificial Intelligence and its Reasonable Application Scenario to Reactor Operation

    International Nuclear Information System (INIS)

    Im, Ki Hong; Suh, Yong-Suk; Park, Cheol; Lim, In-Cheol

    2017-01-01

    This paper presents brief but reasonable scenarios for applying AI or machine learning technologies to research reactor from various perspectives. Two less safety critical scenarios for applying AI to reactor operation are introduced in this study. However, the AI assistant will not only be an assistant but it will also be an operator in the future. What is required is big operation data which can represent all the cases requiring operation decision, including normal operation and accident data as well, and enough time to train and fix the AI system with this data. We can predict AI study in this area can begin with a mild and safe application. But in the near future, this technology could be used to handle or automate more severe operations.

  6. Core design of long life-cycle fast reactors operating without reactivity margin

    International Nuclear Information System (INIS)

    Aristova, E. N.; Baydin, D. F.; Gol'din, V. Y.; Pestryakova, G. A.; Stoynov, M. I.

    2012-01-01

    In this paper we consider a possibility of designing a fast reactor core that operates without reactivity margin for a long time. This study is based on the physical principle of fast reactor operating in a self-adjustable neutron-nuclear regime (SANNR-1) introduced by L.P. Feoktistov (1988-1993) and improved by V. Ya. Gol'din SANNR-2 (1995). The mathematical modeling of active zones of fast reactors in SANNR modes is held by authors since 1992. The numerical simulation is based on solving the neutron transport equation coupled with quasi-diffusion equations. The calculations have been performed using standard 26 energy groups. We use a hierarchy of spatial models of 1D, 1.5D, 2D, and 3D geometries. The spatial models of higher dimensionality are used for verification of results. The calculations showed that operation of the reactor in this mode increases its efficiency, safety and simplifies management. It is possible to achieve continuous work of the reactor in SANNR-2 during 7-10 years without fuel overloads by means of further optimization of the mode. Small reactivity margin is used only for the reactor start up. After first 10-15 days the reactor in SANNR-2 operates without reactivity margin. (authors)

  7. Reactor fuel performance data file, 1985 edition

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Fujita, Misao; Watanabe, Kohji.

    1986-07-01

    In safety evaluation and integrity studies of reactor fuel, data on fuel performance are the most basic materials. The Fuel Reliability Laboratory No.1 has obtained the fuel performance data by joining in some international programs to study the safety and integrity of fuel. Those data have only used for the studies in the above two fields. However, if the data are rearranged and compiled in a easily usable form, they can be utilized in other field of studies. Then, a 'data file' on fuel performance is beeing compiled by adding data from open literatures to those obtained in international programs. The present report is prepared on the basis of the data file compiled by March in 1986. (author)

  8. Development of operator thinking model and its application to nuclear reactor plant operation system

    International Nuclear Information System (INIS)

    Miki, Tetsushi; Endou, Akira; Himeno, Yoshiaki

    1992-01-01

    At first, this paper presents the developing method of an operator thinking model and the outline of the developed model. In next, it describes the nuclear reactor plant operation system which has been developed based on this model. Finally, it has been confirmed that the method described in this paper is very effective in order to construct expert systems which replace the reactor operator's role with AI (artificial intelligence) systems. (author)

  9. Can a nuclear reactor operate for 100 years?

    International Nuclear Information System (INIS)

    Hertel, O.

    2010-01-01

    The TWR (Travelling Wave Reactor) concept was invented in the fifties, then forgotten and it reappeared in 2001 but it was considered too immature to be selected for the fourth generation of nuclear reactors, now an American company 'Terrapower' proposes one whose design is given in the article. This TWR operates with depleted uranium, only the lower part of the fuel rod involves uranium fuel with a civil enrichment ratio (less that 20%). The lower part of the fuel will ignite the fission reaction and enrich the part of fuel just above through neutron absorption. The burning part of the fuel will move up progressively. The main advantage of this reactor is that it can operate for decades without maintenance nor fuel loading. The principle is right on the paper but requires huge technological work to select materials and systems that will be able to withstand decades of operation time in harsh conditions. (A.C.)

  10. Operation of Packed-Bed Reactors Studied in Microgravity

    Science.gov (United States)

    Motil, Brian J.; Balakotaiah, Vemuri

    2004-01-01

    The operation of a packed bed reactor (PBR) involves gas and liquid flowing simultaneously through a fixed-bed of solid particles. Depending on the application, the particles can be various shapes and sizes but are generally designed to force the two fluid phases through a tortuous route of narrow channels connecting the interstitial space. The PBR is the most common type of reactor in industry because it provides for intimate contact and high rates of transport between the phases needed to sustain chemical or biological reactions. The packing may also serve as either a catalyst or as a support for growing biological material. Furthermore, this type of reactor is relatively compact and requires minimal power to operate. This makes it an excellent candidate for unit operations in support of long-duration human space activities.

  11. Considerations of Human Factors in the Design and Operation of Research Reactors

    International Nuclear Information System (INIS)

    Shokr, A.M.

    2015-01-01

    The feedback from the severe accidents occurred at nuclear power plants showed that safety of nuclear installations does not depend only on technical matters but also on human performance. Human errors can initiate an event or can make , by intervention, the event consequences worse. Human factors are of a particular importance for research reactors since the status of these facilities change frequently and the operators have an easy access to the reactor core and to the associated experimental facilities. This paper discusses the experience with human factors and their impact on the safety of research reactors and application of technical and administrative provisions to address these factors in the design and operation phases of research reactors for continuous improvements in safety and performance of these facilities

  12. Control rod for the operation of nuclear reactor

    International Nuclear Information System (INIS)

    Ishida, Hiromi

    1987-01-01

    Purpose: To conduct spectrum shift operation without complicating the reactor core structures, reducing the probability of failures. Constitution: An operation control rod which is driven while passed vertically in the reactor core comprises a strong absorption portion, moderation portion and weak moderation portion defined orderly from above to below and the length for each of the portions is greater than the effective reactor core height. If the operation control rod is lifted to the maximum limit in the upward direction of the reactor core, the weak moderation portion is corresponded over the effective length of the reactor core. Since the weak moderation portion is filled with zirconium and moderators are not present in the operation control rod, water draining gap is formed, neutron spectral shift is formed, excess reactivity is suppressed, absorption of neutrons to fuel fertile material is increased and the formation of nuclear fission material is increased. From the middle to the final stage of the cycle, the control rod is lowered, by which the moderator/fuel effective volume ratio is increased to increase the reactivity. (Kamimura, M.)

  13. Sustainability management for operating organizations of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de, E-mail: ekibrit@ipen.br, E-mail: araquino@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

  14. Sustainability management for operating organizations of research reactors

    International Nuclear Information System (INIS)

    Kibrit, Eduardo; Aquino, Afonso Rodrigues de

    2017-01-01

    Sustainable development is development that meets the needs of the present without compromising the ability of future generations to meet their own needs. In a country like Brazil, where nuclear activity is geared towards peaceful purposes, any operating organization of research reactor should emphasize its commitment to social, environmental, economic and institutional aspects. Social aspects include research and development, production and supply of radiopharmaceuticals, radiation safety and special training for the nuclear sector. Environmental aspects include control of the surroundings and knowledge directed towards environment preservation. Economic aspects include import substitution and diversification of production. Institutional aspects include technology, innovation and knowledge. These aspects, if considered in the management system of an operating organization of research reactor, will help with its long-term maintenance and success in an increasingly competitive market scenario. About this, we propose a sustainability management system approach for operating organizations of research reactors. A bibliographical review on the theme is made. A methodology for identifying indicators for measuring sustainability in nuclear research reactors processes is also described. Finally, we propose a methodology for sustainability perception assessment to be applied at operating organizations of research reactors. (author)

  15. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  16. A review of healthy condition for reactor operators of NPP

    International Nuclear Information System (INIS)

    Zhang Chi; Yang Di; Zhu Lixin; Zhou Limin

    2008-01-01

    Physical health is one of fundamental elements and the necessary qualification for holding operator license or senior operator license of a nuclear power plant issued by NNSA. Based upon requirements for healthy qualification of reactor operators in nuclear safety regulations and some new issued regulations, especially some involved national standard (GB), review direction, relevant contents and criteria were described, and some suggestions were provided in this paper. (authors)

  17. Operation and control of high density tokamak reactors

    International Nuclear Information System (INIS)

    Attenberger, S.E.; McAlees, D.G.

    1976-01-01

    The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

  18. Computer based core monitoring system for an operating CANDU reactor

    International Nuclear Information System (INIS)

    Yoon, Moon Young; Kwon, O Hwan; Kim, Kyung Hwa; Yeom, Choong Sub

    2004-01-01

    The research was performed to develop a CANDU-6 Core Monitoring System(CCMS) that enables operators to have efficient core management by monitoring core power distribution, burnup distribution, and the other important core variables and managing the past core history for Wolsong nuclear power plant unit 1. The CCMS uses Reactor Fueling Simulation Program(RFSP, developed by AECL) for continuous core calculation by integrating the algorithm and assumptions validated and uses the information taken from Digital Control Computer(DCC) for the purpose of producing basic input data. The CCMS has two modules; CCMS server program and CCMS client program. The CCMS server program performs automatic and continuous core calculation and manages overall output controlled by DataBase Management System. The CCMS client program enables users to monitor current and past core status in the predefined GUI(Graphic-User Interface) environment. For the purpose of verifying the effectiveness of CCMS, we compared field-test data with the data used for Wolsong unit 1 operation. In the verification the mean percent differences of both cases were the same(0.008%), which showed that the CCMS could monitor core behaviors well

  19. Operation and maintenance of the RB reactor, Annual report for 1977

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1977-01-01

    The annual report for 1977 includes the following: utilization of the RB reactor; new regulations and instructions for reactor operation; improvement of experimental possibilities of the RB reactor; state of the reactor equipment; dosimetry and radiation protection; reactor staff. Five annexes are concerned with: testing the properties of preamplifiers for linear and logarithmic experimental channels; properties of the neutron converter; maintenance of the reactor equipment; purchase of new equipment; and the program for training reactor operators

  20. Recent performance experience with US light water reactor self-actuating safety and relief valves

    Energy Technology Data Exchange (ETDEWEB)

    Hammer, C.G.

    1996-12-01

    Over the past several years, there have been a number of operating reactor events involving performance of primary and secondary safety and relief valves in U.S. Light Water Reactors. There are several different types of safety and relief valves installed for overpressure protection of various safety systems throughout a typical nuclear power plant. The following discussion is limited to those valves in the reactor coolant systems (RCS) and main steam systems of pressurized water reactors (PWR) and in the RCS of boiling water reactors (BWR), all of which are self-actuating having a setpoint controlled by a spring-loaded disk acting against system fluid pressure. The following discussion relates some of the significant recent experience involving operating reactor events or various testing data. Some of the more unusual and interesting operating events or test data involving some of these designs are included, in addition to some involving a number of similar events and those which have generic applicability.

  1. Intelligence system for reactor operator informational support

    International Nuclear Information System (INIS)

    Prangishvili, I.V.; Pashchenko, F.F.; Saprykin, E.M.

    1989-01-01

    Problems related to creation and introduction at NPP of highly efficient and reliable systems for monitoring and control of working processes and intelligence-endowed systems of operator informational support (ISOIS) are considered. The main units included in ISOIS are considered. The main units included in ISOIS are described. The unit of current state monitoring provides information for the operator, which is necessary under concrete conditions for the process monitoring and control, so as to avoid emergencies and affers a program of actions in a dialogue mode for the operator. The identification unit is designed for the obtaining of assessed values of process parameters (neutron fields, temperatures, pressures) and basic equipment (reactivity coefficients, fuel rod weights, time of delay). The prediction unit evaluates the behaviour of process parameters and process state in various situations. 9 refs

  2. Communication links for fusion reactor maintenance operations

    International Nuclear Information System (INIS)

    Van Uffelen, M.

    2005-01-01

    Different architectures are envisaged for data transmission with fibre optic links in a radiation environment, as proposed in literature for both space and high energy physics applications. Their needs and constraints differ from those encountered for maintenance tasks in the future ITER environment, not only in terms of temperature and radiation levels, but also with respect to transmission speed requirements. Our approach attempts to limit the use of radiation-sensitive electronics for transmission of both digital and/or analogue data to the control room, using glass fibres as transport medium. We therefore assessed the radiation behaviour of a cost-effective fibre optic transmitter at 850 nm, consisting of a PWM (pulse width modulator), a radiation tolerant current driver (previously developed at SCK-CEN) and a VCSEL (Vertical-Cavity Surface Emitting Laser assembly, up to 10 MGy at 60 degrees Celsius. The PWM enables to transform an analogue sensor signal into a pseudo numerical signal, with a pulse width proportional to the incoming signal. The main objective of this task is to contribute to the major design of the maintenance equipment and strategy needed for the remote replacement of the divertor system in the future ITER fusion reactor, with particular attention to the implications of radiation hardening rules and recommendations. Next to the radiation assessment studies of remote handling tools, including actuators and sensors, we also develop radiation tolerant communication links with multiplexing capabilities

  3. AMNT 2014. Key topic: Reactor operation, safety - report. Pt. 3

    Energy Technology Data Exchange (ETDEWEB)

    Bohnstedt, Angelika [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany). Programm Nukleare Sicherheitsforschung (NUKLEAR); Mull, Thomas [AREVA GmbH, Erlangen (Germany). Nuclear Fusion, HTR and Transverse Issues (PTDH-G); Starflinger, Joerg [Stuttgart Univ. (Germany). Inst. fuer Kernenergetik und Energiesysteme (IKE)

    2015-01-15

    Summary report on the following sessions of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Reactor Operation, Safety: Radiation Protection (Angelika Bohnstedt); - Competence, Innovation, Regulation: Fusion Technology - Optimisation Steps in the ITER Design (Thomas Mull); - Competence, Innovation, Regulation: Education, Expert Knowledge, Knowledge Transfer (Joerg Starflinger). The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' have been covered in atw 10 and 12 (2015) and will be covered in further issues of atw.

  4. Method of operating water cooled reactor with blanket

    International Nuclear Information System (INIS)

    Suzuki, Katsuo.

    1988-01-01

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

  5. Jules Horowitz Reactor: Organisation for the Preparation of the Commissioning Phase and Normal Operation

    Energy Technology Data Exchange (ETDEWEB)

    Estrade, J.; Fabre, J. L.; Marcille, O. [French Alternative Energies end Atomic Energy Commission, Provence (France)

    2013-07-01

    The Jules Horowitz Reactor (JHR) is a new modern Material Testing Reactor (MTR) currently under construction at CEA Cadarache research centre in the south of France. It will be a major research facility in support to the development and the qualification of materials and fuels under irradiation with sizes and environment conditions relevant for nuclear power plants in order to optimise and demonstrate safe operations of existing power reactors as well as to support future reactors design. It will represent also an important research infrastructure for scientific studies dealing with material and fuel behaviour under irradiation. The JHR will contribute also to secure the production of radioisotope for medical application. This is a key public health stake. The construction of JHR which started in 2007 is going-on with target of commissioning by the end of 2017. The design of the reactor provides modern experimental capacity in support to R and D programs for the nuclear energy for the next 60 years. In parallel to the facility construction, the preparation of the future staff and of the organisation to operate the reactor safely, reliably and efficiently is an important issue. In this framework, many actions are in progress to elaborate: Ο the staffing and the organisational structure for the commissioning test phases and also for normal operation, Ο the documentation in support to the reactor operation (safety analysis report, general operating rules, procedures, instructions, ···), Ο the maintenance, in service and periodic test programs, Ο staff training programs by using dedicated facilities (simulator, ···) Ο commissioning test programs for ensuring that the layout of systems and subcomponents is completed in accordance with the design requirements, the specification performances and the safety criteria. These commissioning tests will also be helpful for transferring the knowledge on the installed systems to the operating group. This paper gives the

  6. The need for high performance breeder reactors

    International Nuclear Information System (INIS)

    Vaughan, R.D.; Chermanne, J.

    1977-01-01

    It can be easily demonstrated, on the basis of realistic estimates of continued high oil costs, that an increasing portion of the growth in energy demand must be supplied by nuclear power and that this one might account for 20% of all the energy production by the end of the century. Such assumptions lead very quickly to the conclusion that the discovery, extraction and processing of the uranium will not be able to follow the demand; the bottleneck will essentially be related to the rate at which the ore can be discovered and extracted, and not to the existing quantities nor their grade. Figures as high as 150.000 T/annum and more would be quickly reached, and it is necessary to wonder already now if enough capital can be attracted to meet these requirements. There is only one solution to this problem: improve the conversion ratio of the nuclear system and quickly reach the breeding; this would lead to the reduction of the natural uranium consumption by a factor of about 50. However, this condition is not sufficient; the commercial breeder must have a breeding gain as high as possible because the Pu out-of-pile time and the Pu losses in the cycle could lead to an unacceptable doubling time for the system, if the breeding gain is too low. That is the reason why it is vital to develop high performance breeder reactors. The present paper indicates how the Gas-cooled Breeder Reactor [GBR] can meet the problems mentioned above, on the basis of recent and realistic studies. It briefly describes the present status of GBR development, from the predecessors in the gas cooled reactor line, particularly the AGR. It shows how the GBR fuel takes mostly profit from the LMFBR fuel irradiation experience. It compares the GBR performance on a consistent basis with that of the LMFBR. The GBR capital and fuel cycle costs are compared with those of thermal and fast reactors respectively. The conclusion is, based on a cost-benefit study, that the GBR must be quickly developed in order

  7. Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

    International Nuclear Information System (INIS)

    Abou Elmaaty, T.

    2014-01-01

    The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the suing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer is implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer is produced an inverse buoyant force make the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow - open pool research reactor (with a power greater than 20 M watt) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against Gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability

  8. Experimental Investigation of the Hot Water Layer Effect on Upward Flow Open Pool Reactor Operability

    International Nuclear Information System (INIS)

    Abou Elmaaty, T.

    2015-01-01

    The open pool reactor offers a high degree of reliability in the handling and manoeuvring, the replacement of reactor internal components and the swing of vertical irradiation channels. The protection of both the operators and the reactor hall environment against radiation hazards is considered a matter of interest. So, a hot water layer implemented above many of the research reactors main pool, especially those whose flow direction is upward flow. An experimental work was carried out to ensure the operability of the upward flow open pool research reactor with / without the hot water layer. The performed experiment showed that, the hot water layer produced an inverse buoyant force making the water to diffuse downward against the ordinary natural circulation from the reactor core. An upward flow-open pool research reactor (with a power greater than 20 Mw) could not wok without a hot water layer. The high temperature of the hot water layer surface could release a considerable amount of water vapour into the reactor hall, so a heat and mass transfer model is built based on the measured hot water layer surface temperature to calculate the amount of released water vapour during the reactor operating period. The effects of many parameters like the ambient air temperature, the reactor hall relative humidity and the speed of the pushed air layer above the top pool end on the evaporation rate is studied. The current study showed that, the hot water layer system is considered an efficient shielding system against gamma radiation for open pool upward flow reactor and that system should be operated before the reactor start up by a suitable period of time. While, the heat and mass transfer model results showed that, the amount of the released water vapour is increased as a result of both the increase in hot water layer surface temperature and the increase in air layer speed. As the increase in hot water layer surface temperature could produce a good operability conditions from

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

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

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

  10. Performance of Fragema fuel in pressurized water reactors

    International Nuclear Information System (INIS)

    Dumont, A.; Ravier, G.; Ballot, B.

    1986-06-01

    FRAGEMA fuel operating experience in power reactors is very extensive. Performance over a range of power and burnup levels for various operating conditions is quite satisfactory. However significant development programs are presently in progress to further extend our knowledge under increasingly severe operating conditions. In particular, upcoming data acquisition programs (1985-1988) will cover site and hot cell measurements on Gd poison rods, 4.5 % overenriched fuel rods over four operating cycles, 17 x 17 AFA fuel assemblies. For these products the same surveillance strategy as the one used for the standard assembly has been adopted, in order to continuously provide more data which can be used to upgrade design models and pave the way for the development of future products

  11. Operating history report for the Peach Bottom HTGR. Volume I. Reactor operating history

    International Nuclear Information System (INIS)

    Scheffel, W.J.; Baldwin, N.L.; Tomlin, R.W.

    1976-01-01

    The operating history for the Peach Bottom-1 Reactor is presented for the years 1966 through 1975. Information concerning general chemistry data, general physics data, location of sensing elements in the primary helium circuit, and postirradiation examination and testing of reactor components is presented

  12. Computer codes for the operational control of the research reactors

    International Nuclear Information System (INIS)

    Kalker, K.J.; Nabbi, R.; Bormann, H.J.

    1986-01-01

    Four small computer codes developed by ZFR are presented, which have been used for several years during operation of the research reactors FRJ-1, FRJ-2, AVR (all in Juelich) and DR-2 (Riso, Denmark). Because of interest coming from the other reactor stations the codes are documented within the frame work of the IAEA Research Contract No. 3634/FG. The zero-dimensional burnup program CREMAT is used for reactor cores in which flux measurements at each individual fuel element are carried out during operation. The program yields burnup data for each fuel element and for the whole core. On the basis of these data, fuel reloading is prepared for the next operational period under consideration of the permitted minimum shut down reactivity of the system. The program BURNY calculates burnup for fuel elements inaccessible for flux measurements, but for which 'position weighting factors' have been measured/calculated during zero power operation of the core, and which are assumed to be constant in all operational situations. The code CURIAX calculates post-irradiation data for discharged fuel elements needed in their manipulation and transport. These three programs have been written for highly enriched fuel and take into account U-235 only. The modification of CREMAT for LEU Cores and its combiantion with ORIGEN is in preparation. KINIK is an inverse kinetic code and widely used for absorber rod calibration at the abovementioned research reactors. It includes a special polynomial subroutine which can easily be used in other codes. (orig.) [de

  13. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

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

  14. RHIC heavy ion operations performance

    CERN Document Server

    Satogata, T; Ferrone, R; Pilat, F

    2006-01-01

    The Relativistic Heavy Ion Collider (RHIC) completed its fifth year of operation in 2005, colliding copper ion beams with ps=200 GeV/u and 62.4 GeV/u[1]. Previous heavy ion runs have collided gold ions at ps=130 GeV/u, 200 GeV/u, and 62.4 GeV/u[2], and deuterons and gold ions at ps=200 GeV/u[3]. This paper discusses operational performance statistics of this facility, including Cu- Cu delivered luminosity, availability, calendar time spent in physics stores, and time between physics stores. We summarize the major factors affecting operations efficiency, and characterize machine activities between physics stores.

  15. Graphics and control for in-reactor operations

    International Nuclear Information System (INIS)

    Smith, A.L.

    1996-01-01

    A wide range of manipulator systems has been developed to carry out remotely operated inspection, repair and maintenance tasks at the Magnox reactors in the United Kingdom. A key factor in the improvement of these systems in recent years has been the extensive use of computer graphics as a real-time aid to the manipulator operator. This is exemplified by the reactor pressure vessel inspection work at the Bradwell reactor which is described in detail. The graphics sub-system of the control system for the manipulator plays a unique and wide-ranging role. The 3D modelling and simulation capability of the IGRIP software has contributed to the conceptual design, detailed path planning, rehearsal support, public relations, real-time manipulator display, post inspection documentation and quality assurance. (UK)

  16. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    1989-06-01

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  17. Operation of the BR2 Reactor

    International Nuclear Information System (INIS)

    Gubel, P.

    2006-01-01

    The BR2 is still SCK-CEN's most important nuclear facility. After an extensive refurbishment of 22 months to compensate for the ageing of the installations, to enhance the reliability of operation and to comply with modern safety standards, it was restarted in April 1997. The facility is mainly used for the irradiation and testing of fuels and materials and for commercial productions - including radioisotopes for the medical and industrial uses, and NTD-Silicon. The article describes the main achievements and activities in 2005

  18. IAEA support for operating nuclear reactors

    International Nuclear Information System (INIS)

    Akira, O.

    2010-01-01

    The IAEA programme, under the pillar of science and technology, provides support to the existing fleet of nuclear power plants (NPPs) for excellence in operation, support to new countries for infrastructure development, stimulating technology innovation for sustainable development and building national capability. Practical activities include methodology development, information sharing and providing guidance documents and state-of-the-art reports, networking of research activities, and review services using guidance documents as a basis of evaluation. This paper elaborates more on the IAEA's activities in support of the existing fleet of nuclear power plants

  19. Operation of the BR2 Reactor

    International Nuclear Information System (INIS)

    Gubel, P.

    2005-01-01

    The BR2 is still SCK-CEN's most important nuclear facility. After an extensive refurbishment of 22 months to compensate for the ageing of the installations, to enhance the reliability of operation and to comply with modern safety standards, it was restarted in April 1997. The facility is mainly used for the irradiation and testing of fuels and materials and for commercial productions - including radioisotopes for the medical and industrial uses, and NTD-Silicon. The article describes the main activities and achievements in 2004

  20. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    International Nuclear Information System (INIS)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao

    1997-11-01

    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship 'Mutsu'. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

  1. Development of advanced automatic control system for nuclear ship. 2. Perfect automatic operation after reactor scram events

    Energy Technology Data Exchange (ETDEWEB)

    Yabuuchi, Noriaki; Nakazawa, Toshio; Takahashi, Hiroki; Shimazaki, Junya; Hoshi, Tsutao [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    An automatic operation system has been developed for the purpose of realizing a perfect automatic plant operation after reactor scram events. The goal of the automatic operation after a reactor scram event is to bring the reactor hot stand-by condition automatically. The basic functions of this system are as follows; to monitor actions of the equipments of safety actions after a reactor scram, to control necessary control equipments to bring a reactor to a hot stand-by condition automatically, and to energize a decay heat removal system. The performance evaluation on this system was carried out by comparing the results using to Nuclear Ship Engineering Simulation System (NESSY) and the those measured in the scram test of the nuclear ship `Mutsu`. As the result, it was showed that this system had the sufficient performance to bring a reactor to a hot syand-by condition quickly and safety. (author)

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

    International Nuclear Information System (INIS)

    Ho, V.; Apostolakis, G.

    1988-01-01

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

  3. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    2007-12-01

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  4. Approach to training the operators of WWER-440 reactors

    International Nuclear Information System (INIS)

    Pironkov, L.; Minakova, R.

    2002-01-01

    The paper has three parts. (1) Personnel Training and Qualifications (2) Description of Kozloduy NPP Training and Qualification System (TQS) built in the last 7 years and its interfaces with the Certification System and (3) Application of the TQS for the Senior Reactor Operator (SRO). (author)

  5. 1982 Annual status report: operation of the high flux reactor

    International Nuclear Information System (INIS)

    1983-01-01

    The high flux materials testing reactor has been operated in 1982 within a few percent of the pre-set schedule, attaining 73% overall availability. Its utilization reached another record figure in 20 years: 81% without, 92% with, the low enrichment test elements irradiated during the year

  6. Best Safety Practices for the Operation of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Boeck, H.; Villa, M. [Atominstitute of the Austrian Universities, 1020 Vienna (Austria)

    2002-07-01

    A survey on administrative, organisational and technical aspects for the safe and efficient operation of a 250 kW TRIGA Mark II research reactor is given. The replacement of the I and C system is discussed, maintenance procedures are presented and the fuel management is described. (author)

  7. AMNT 2014. Key Topic: Reactor operation, safety - report. Pt. 1

    International Nuclear Information System (INIS)

    Schaffrath, Andreas

    2014-01-01

    Summary report on one session of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Safety of Nuclear Installations - Methods, Analysis, Results: Backfittings for the Improvement of Safety and Efficiency. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' will be covered in further issues of atw.

  8. Periodic Operation of Three-Phase Catalytic Reactors

    Czech Academy of Sciences Publication Activity Database

    Silveston, P.T.; Hanika, Jiří

    2005-01-01

    Roč. 82, č. 6 (2005), s. 1105-1142 ISSN 0008-4034 Institutional research plan: CEZ:AV0Z4072921 Keywords : three-phase reactors * trickle bed * periodic operation Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 0.574, year: 2005

  9. Best Safety Practices for the Operation of Research Reactors

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2002-01-01

    A survey on administrative, organisational and technical aspects for the safe and efficient operation of a 250 kW TRIGA Mark II research reactor is given. The replacement of the I and C system is discussed, maintenance procedures are presented and the fuel management is described. (author)

  10. Computerised programming of the Dragon reactor fuel handling operations

    International Nuclear Information System (INIS)

    Butcher, P.

    1976-11-01

    Two suites of FORTRAN IV computer programs have been written to produce check lists for the operation of the two remote control fuel handling machines of the Dragon Reactor. This document describes the advantages of these programs over the previous manual system of writing check lists, and provides a detailed guide to the programs themselves. (author)

  11. Operational readiness reviews for restart of L reactor

    International Nuclear Information System (INIS)

    Finley, R.H.

    1984-01-01

    The L Reactor at the Savannah River Plant is being restarted after being in a standby status since 1968. Operational Readiness Reviews (ORRs) were conducted by DOE-SR and contractor personnel concurrent with the restart activity. This paper summarizes the ORR activity

  12. Groundwater Monitoring Plan for the Reactor Technology Complex Operable Unit 2-13

    International Nuclear Information System (INIS)

    Richard P. Wells

    2007-01-01

    This Groundwater Monitoring Plan describes the objectives, activities, and assessments that will be performed to support the on-going groundwater monitoring requirements at the Reactor Technology Complex, formerly the Test Reactor Area (TRA). The requirements for groundwater monitoring were stipulated in the Final Record of Decision for Test Reactor Area, Operable Unit 2-13, signed in December 1997. The monitoring requirements were modified by the First Five-Year Review Report for the Test Reactor Area, Operable Unit 2-13, at the Idaho National Engineering and Environmental Laboratory to focus on those contaminants of concern that warrant continued surveillance, including chromium, tritium, strontium-90, and cobalt-60. Based upon recommendations provided in the Annual Groundwater Monitoring Status Report for 2006, the groundwater monitoring frequency was reduced to annually from twice a year

  13. Thermal-structural response of EBR-II major components under reactor operational transients

    International Nuclear Information System (INIS)

    Chang, L.K.; Lee, M.J.

    1983-01-01

    Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

  14. Training and qualification of licensed reactor operators at General Public Utilities Nuclear Corporation

    International Nuclear Information System (INIS)

    Long, R.L.; Coe, R.P.

    1992-01-01

    Following the Three Mile Island-2 (TMI-2) accident in 1979, the utility responsible for managing the facility has looked closely at the training and qualification of its reactor operators. Performance-based operator training programmes are now in place, as required by the United States National Academy for Nuclear Training. Operators also participate directly in the development of a professional code of behaviour. (UK)

  15. Reactor operating procedures for start up of continuously operated chemical plants

    NARCIS (Netherlands)

    Verwijs, J.W.; Verwijs, J.W.; Kösters, P.H.; van den Berg, Henderikus; Westerterp, K.R.; Kosters, P.G.H.

    1995-01-01

    Rules are presented for the startup of an adiabatic tubular reactor, based on a qualitative analysis of the dynamic behavior of continuously-operated vapor- and liquid-phase processes. The relationships between the process dynamics, operating criteria, and operating constraints are investigated,

  16. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  17. Annual progress report 1988, operation of the high flux reactor

    International Nuclear Information System (INIS)

    1989-01-01

    In 1988 the High Flux Reactor Petten was routinely operated without any unforeseen event. The availability was 99% of scheduled operation. Utilization of the irradiation positions amounted to 80% of the practical occupation limit. The exploitation pattern comprised nuclear energy deployment, fundamental research with neutrons, and radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

  18. Designing visual displays and system models for safe reactor operations

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.

    1995-01-01

    The material presented in this paper is based on two studies involving the design of visual displays and the user's prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator's perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors

  19. 1980 Annual status report: operation of the high flux reactor

    International Nuclear Information System (INIS)

    1981-01-01

    HFR Petten has been operated in 1980 in fulfilment of the 1980/83 JRC Programme Decision. Both reactor operation and utilization data have been met within a few percent of the goals set out in the annual working schedule, in support of a large variety of research programmes. Major improvements to experimental facilities have been introduced during the year and future modernization has been prepared

  20. Introduction of long term cycle of reactor operation

    International Nuclear Information System (INIS)

    Aoyati, M.; Tanaka, T.

    2004-01-01

    Introduction of long term cycle of LWR reactor operation at NPP in Japan is considered, and problems of technical, legislative and economical character, increase of power coefficient are discussed. More long term operation period provides decreasing frequency of periodic inspections and reduction of personnel radiation doses. Reliability of fuel, energetic equipment, mechanisms and devices must be taken into account for the decision of technical problems. Consumptions for electric power generation are studied [ru

  1. I and C system at TRIGA - ICN reactor after more than 20 years of operation

    International Nuclear Information System (INIS)

    Ionila, M.; Preda, M.

    2002-01-01

    An I and C system that is involved in a nuclear safety function has to be itself safe in operation, strictly performing the survey of those parameters, which are linked to the safety function. The precision of such a system is sufficient for the safety function, but for a more accurate evaluation of the in-core experimental phenomena, the presence of a data acquisition and processing system is needed. The two systems must be together taken into account by the reactor operation. The data acquisition and processing system designated for the monitoring of the stationary or the slow-varying processes allow the safety function evaluation from the point of view of the statistics of the effective reactor operation time along a certain period of time. The evaluation of the unscheduled reactor shutdowns determined by those reactor systems having safety functions with the percentage contribution of each system is presented. The data were selected from the annual operation reports for the reactor and the reactor installations in the period 1981-1999

  2. Design and performance of subgrade biogeochemical reactors.

    Science.gov (United States)

    Gamlin, Jeff; Downey, Doug; Shearer, Brad; Favara, Paul

    2017-12-15

    Subgrade biogeochemical reactors (SBGRs), also commonly referred to as in situ bioreactors, are a unique technology for treatment of contaminant source areas and groundwater plume hot spots. SBGRs have most commonly been configured for enhanced reductive dechlorination (ERD) applications for chlorinated solvent treatment. However, they have also been designed for other contaminant classes using alternative treatment media. The SBGR technology typically consists of removal of contaminated soil via excavation or large-diameter augers, and backfill of the soil void with gravel and treatment amendments tailored to the target contaminant(s). In most cases SBGRs include installation of infiltration piping and a low-flow pumping system (typically solar-powered) to recirculate contaminated groundwater through the SBGR for treatment. SBGRs have been constructed in multiple configurations, including designs capable of meeting limited access restrictions at heavily industrialized sites, and at sites with restrictions on surface disturbance due to sensitive species or habitat issues. Typical performance results for ERD applications include 85 to 90 percent total molar reduction of chlorinated volatile organic compounds (CVOCs) near the SBGR and rapid clean-up of adjacent dissolved contaminant source areas. Based on a review of the literature and CH2M's field-scale results from over a dozen SBGRs with a least one year of performance data, important site-specific design considerations include: 1) hydraulic residence time should be long enough for sufficient treatment but not too long to create depressed pH and stagnant conditions (e.g., typically between 10 and 60 days), 2) reactor material should balance appropriate organic mulch as optimal bacterial growth media along with other organic additives that provide bioavailable organic carbon, 3) a variety of native bacteria are important to the treatment process, and 4) biologically mediated generation of iron sulfides along with

  3. RA Research reactor Annual report 1982 - Part 1, Operation, maintenance and utilization of the RA reactor

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Radivojevic, J.; Stamenkovic, D.; Skoric, M.; Miokovic, J.

    1982-12-01

    Reactor test operation started in September 1981 at 2 MW power with 80% enriched fuel continued during 1982 according to the previous plan. The initial reactor core was made of 44 fuel channel each containing 10 fuel slugs. The first half of 1982 was used for the needed measurements and analysis of operating parameters and functioning of reactor systems and equipment under operating conditions. Program concerned with the testing operation at higher power levels was started in the second half of this year. It was found that the inherent excess reactivity and control rod worths ensure safe operation according to the IAEA safety standards. Excess reactivity is high enough to enable higher power level of 4.7 MW during 4 monthly cycles each lasting 15-20 days. Favourable conditions for cooling exist for the initial core configuration. Effects of poisoning at startup on the reactivity and power density distribution were measured as well as initial spatial distribution of the neutron flux which was 3,9 10 13 cm -2 s -1 at 2 MW power. Modification of the calibration coefficient in the system for automated power level control was determined. All the results show that all the safety criteria and limitations concerned with fuel utilization are fulfilled if reactor power would be 4.7 MW. Additional testing operation at 3, 4, and 4.7 MW power levels will be needed after obtaining the licence for operating at nominal power. Transition from the initial core with 44 fuel channels to the equilibrium lattice configuration with 72 fuel channels each containing 10 fuel slugs, would be done gradually. Reactor was not operated in September because of the secondary coolant pipes were exchanged between Danube and the horizontal sedimentary. Control and maintenance of the reactor equipment was done regularly and efficiently dependent on the availability of the spare parts. Difficulties in maintenance of the reactor instrumentation were caused by unavailability of the outdated spare parts

  4. Regulation concerning installation and operation of reactors for power generation

    International Nuclear Information System (INIS)

    1987-01-01

    This report shows the Ordinance of the Ministry of International Trade and Industry No.77 of December 28, 1978. The ordinance consists of provisions covering application for permission for construction of nuclear reactor (concerning continuous maximum thermal output, location, structure, reactor core, fuel material, moderator, reflector, cooling system, measurement control system, safety circuit, control system, emergency system, radioactive waste proposal facilities, construction plan, meteorology and other environmental conditions, etc.), operation plan (to be submitted every year), application for approval of joint management (name, address, facilities location, conditions for joint management, etc.), cancellation of permission (in five years from the date of permission), record keeping (density and temperature of neutron, temperature and pressure of coolant, purity of mederator, etc.), restriction on access to areas under management (measures to be taken in such areas), measures concerning exposure to radioactive rays (allowable dosage, etc.), patrol and checking in nuclear reactor facilities, self-imposed regular inspection of nuclear reactor facilities, operation of nuclear reactor, transport within plant or business establishment, storage (storing facilities, etc.), waste disposal, etc. (Nogami, K.)

  5. Steady-state operation requirements of tokamak fusion reactor concepts

    International Nuclear Information System (INIS)

    Knobloch, A.F.

    1991-06-01

    In the last two decades tokamak conceptual reactor design studies have been deriving benefit from progressing plasma physics experiments, more depth in theory and increasing detail in technology and engineering. Recent full-scale reactor extrapolations such as the US ARIES-I and the EC Reference Reactor study provide information on rather advanced concepts that are called for when economic boundary conditions are imposed. The ITER international reactor design activity concentrated on defining the next step after the JET generation of experiments. For steady-state operation as required for any future commercial tokamak fusion power plants it is essential to have non-inductive current drive. The current drive power and other internal power requirements specific to magnetic confinement fusion have to be kept as low as possible in order to attain a competitive overall power conversion efficiency. A high plasma Q is primarily dependent on a high current drive efficiency. Since such conditions have not yet been attained in practice, the present situation and the degree of further development required are characterized. Such development and an appropriately designed next-step tokamak reactor make the gradual realization of high-Q operation appear feasible. (orig.)

  6. 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)

  7. Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Rokhmadi

    2007-01-01

    Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

  8. Control of operational transients in power reactors - Methodology

    International Nuclear Information System (INIS)

    Vukovic, D.

    1983-01-01

    By introducing the nuclear power stations in the electric power system, questions of their possibilities to satisfy system's demand arise. Control of operational transients (temperature and Xe 135 ) in power reactors by determining the optimal control rod strategy is given. Ti optimize the Xe 135 transients, the Pantryagin theorem of optimal processes is applied. For solving three dimensional, two-group diffusion equations the heterogeneous Feinberg-Galanin method with axial flux harmonics is adopted. An application of this formalism to three-dimensional, finite cylindrical pressurised water reactor radially reflected is presented. (author)

  9. Measurements of operator performance - an experimental setup

    International Nuclear Information System (INIS)

    Netland, K.

    1980-01-01

    The human has to be considered as an important element in a process control system, even if the degree of automation is extremely high. Other elements, e.g. computer, displays, etc., can to a large extent be described and quantified. The human (operator), is difficult to describe in a precise way, and it is just as difficult to predict his thinking and acting in a control room environment. Many factors influence his performance, such as: experience, motivation, level of knowledge, training, control environment, job organization, etc. These factors have to a certain degree to be described before guidelines for design of the man-process interfaces and the control room layout can be developed. For decades, the psychological science has obtained knowledge of the human mind and behaviour. This knowledge should have the potential of a positive input on our effort to describe the factors influencing the operator performance. Even if the human is complex, a better understanding of his thinking and acting, and a more precise description of the factors influencing his performance can be obtained. At OECD Halden Reactor Project an experimental set-up for such studies has been developed and implemented in the computer laboratory. The present set-up includes elements as a computer- and display-based control room, a simulator representing a nuclear power plant, training programme for the subjects, and methods for the experiments. Set-up modules allow reconfiguration of experiments. (orig./HP)

  10. Operation and maintenance of the RB reactor, Annual report for 1980

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.; Markovic, H.; Zivkovic, B.; Gogdanovic, M.; Petronijevic, M.

    1980-12-01

    This report includes data concerned with reactor operation and utilization, status of reactor components and equipment, refurbishment of the equipment, dosimetry and radiation protection, reactor staff, financing. It includes 9 Annexes as follows: Utilization of the RB reactor from 1976 - 1980; program of reactor utilization from 1981-1985; contents of the RB reactor safety report; maintenance of the reactor components and equipment in 1980; verification of reactor reliability after the earthquake (May 18 1980); refurbishment of equipment in 1980, and purchasing new equipment from 1981-1985; review of radiation doses in the reactor building and exposure doses for the reactor staff; personnel data and financial data

  11. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  12. Radioactive effluents from CANDU 6 reactors during normal operation

    International Nuclear Information System (INIS)

    Boss, C.R.; Allsop, P.J.

    1995-12-01

    During routine operation of a CANDU 6 reactor, various gaseous, liquid, and solid radioactive wastes are generated. The layout of the CANDU 6 reactor and the design of its systems ensure that these are minimized, but small quantities of gaseous and liquid wastes are continually discharged at very low concentrations. This report discusses the make-up of these chronically generated gaseous and liquid effluents. From a safety perspective, the doses to individual members of the public resulting from radioactive wastes chronically discharged from CANDU 6 reactors have been negligible. Similarly, doses to the regional and global populations have been negligible, generally less than 0.001% of background. While far below regulatory limits, releases of tritium, noble gases and gross β - -γ have been the most radiologically significant emissions, while radioiodine and particulates have had the greatest potential to deliver public dose. (author). 8 refs., 16 tabs., 3 figs

  13. Safeguards operations in the integral fast reactor fuel cycle

    International Nuclear Information System (INIS)

    Goff, K.M.; Benedict, R.W.; Brumbach, S.B.; Dickerman, C.E.; Tompot, R.W.

    1994-01-01

    Argonne National Laboratory is currently demonstrating the fuel cycle for the Integral Fast Reactor (IFR), an advanced reactor concept that takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel-cycle economics, environmental protection, and safeguards. The IFR fuel cycle employs a pyrometallurgical process using molten salts and liquid metals to recover actinides from spent fuel. The safeguards aspects of the fuel cycle demonstration must be approved by the United States Department of Energy, but a further goal of the program is to develop a safeguards system that could gain acceptance from the Nuclear Regulatory Commission and International Atomic Energy Agency. This fuel cycle is described with emphasis on aspects that differ from aqueous reprocessing and on its improved safeguardability due to decreased attractiveness and diversion potential of all process streams, including the fuel product

  14. Diamond Ordinance Radiation Facility (DORF) reactor operating experiences

    International Nuclear Information System (INIS)

    Gieseler, Walter

    1970-01-01

    The Diamond Ordnance Radiation Facility Mark F Reactor is described and some of the problems encountered with its operation are discussed. In a period from reactor startup in September 1961 to June 1964, when the aluminum-clad core was changed to a stainless-steel clad core, a total of 30 fuel elements were removed from reactor service because of excessive growth. One leaking fuel element was detected during the lifetime of the aluminum- clad core. In June 1964, the core was changed to the stainless-steel-clad high hydride fuel elements. Since the installation of the stainless-steel-clad fuel element core, there has been a gradual decline of excess reactivity. Various theories were discussed as the cause but the investigations have resulted in no definitive conclusion that could account for the total reactivity loss

  15. Improvements in operational safety performance of the Magnox power stations

    Energy Technology Data Exchange (ETDEWEB)

    Marchese, C.J. [BNFL Magnox Generation, Berkeley (United Kingdom)

    2000-10-01

    In the 43 years since commencement of operation of Calder Hall, the first Magnox power station, there remain eight Magnox stations and 20 reactors still in operation, owned by BNFL Magnox Generation. This paper describes how the operational safety performance of these stations has significantly improved over the last ten years. This has been achieved against a background of commercial competition introduced by privatization and despite the fact that the Magnox base design belongs to the past. Finally, the company's future plans for continued improvements in operational safety performance are discussed. (author)

  16. Safe design and operation of fluidized-bed reactors: Choice between reactor models

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1990-01-01

    For three different catalytic fluidized bed reactor models, two models presented by Werther and a model presented by van Deemter, the region of safe and unique operation for a chosen reaction system was investigated. Three reaction systems were used: the oxidation of benzene to maleic anhydride, the

  17. Evaluation of the revised electrolytic reduction reactor from a remote operability aspect

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyo Jik; Yoon, K. H.; Kim, K. H. (and others)

    2008-01-15

    This report presents an evaluation of the remote operability of the revised electrolytic reduction (ER) reactor installed in the ACP at KAERI. All operations have to be implemented in a fully remote manner since the ACPF is a hotcell for handling highly radioactive materials such as spent nuclear fuels. In particular, the ER process is a key process of the Advanced spent fuel Conditioning Process (ACP) and it needs a lot of other auxiliary equipment to perform it. Also, since the ER equipment is too big and complicated, and contrarily the ACPF is not big enough, and one common rail is allotted for a bridge transported servo manipulator, an in-cell crane and a gate crane, a remote handling of the ER reactor is accompanied by a difficulty for a remote operation. For an easy understanding, short overview of the ER process, the remote handling equipment, the structural configuration of the ACPF and the detail drawings of the ER equipment are presented. Through 4 month-remote operational tests, detailed operational procedures are presented along with pictures. The remote handling equipment and tools required in each operation are addressed in detail. Also, the procedure to implement each remote operation, and a task difficulty are evaluated from a remote operability aspect. All the remote tasks are distinguished according to whether a remote operation can be performed or not. Finally, partial improvement or an idea to solve the suggested problems is presented. This report will assist in modifying or scaling up the ER reactor.

  18. Fuzzy operation and real time surveillance of a nuclear reactor

    International Nuclear Information System (INIS)

    Si-Fodil, M.; Guely, F.; Siarry, P.; Tyran, J.L.

    1997-01-01

    The operating power of nuclear power plants needs to be modulated according to the thin evolutions of electric power demand. Two parameters are concerned by load following operations: the power axial disequilibrium and the position of control rods. This paper deals with the automation of the control of power axial disequilibrium using boration-dilution. An automatic system based on fuzzy logic is proposed which can be substituted to the expert operator who is in charge of this fastidious manual task. The management of water and boron flow rates are studied in details. A Graphic interface was designed for the real-time surveillance of the reactor. (J.S.)

  19. Meeting the reactor operator's information needs using functional analysis

    International Nuclear Information System (INIS)

    Nelson, W.R.; Clark, M.T.

    1980-01-01

    Since the accident at Three Mile Island, many ideas have been proposed for assisting the reactor operator during emergency situations. However, some of the suggested remedies do not alleviate an important shortcoming of the TMI control room: the operators were not presented with the information they needed in a manner which would allow prompt diagnosis of the problem. To address this problem, functional analysis is being applied at the LOFT facility to ensure that the operator's information needs are being met in his procedures and graphic displays. This paper summarizes the current applications of functional analysis at LOFT

  20. Plasma engineering analyses of tokamak reactor operating space

    International Nuclear Information System (INIS)

    Houlberg, W.; Attenberger, S.E.

    1981-01-01

    A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models

  1. Calculations of radiation levels during reactor operations for safeguard inspections

    International Nuclear Information System (INIS)

    Sobhy, M.

    1996-01-01

    When an internal core spent fuel storage is used in the shield tank to accommodate a large number of spent fuel baskets, physical calculations are performed to determine the number of these spent fuel elements which can be accommodated and still maintain the gamma activity outside under the permissible limit. The corresponding reactor power level is determined. The radioactivity calculations are performed for this internal storage at different axial levels to avoid the criticality of the reactor core. Transport theory is used in calculations based on collision probability for multi group cell calculations. Diffusion theory is used in three dimensions in the core calculations. The nuclear fuel history is traced and radioactive decay is calculated, since reactor fission products are very sensitive to power level. The radioactivity is calculated with a developed formula based on fuel basket loading integrity. (author)

  2. Microorganism selection and biosurfactant production in a continuously and periodically operated bioslurry reactor.

    Science.gov (United States)

    Cassidy, D P; Hudak, A J

    2001-06-29

    A continuous-flow reactor (CSTR) and a soil slurry-sequencing batch reactor (SS-SBR) were maintained in 8l vessels for 180 days to treat a soil contaminated with diesel fuel (DF). Concentrations of Candida tropicalis, Brevibacterium casei, Flavobacterium aquatile, Pseudomonas aeruginosa, and Pseudomonas fluorescens were determined using fatty acid methyl ester (FAME) analysis. DF removal (biological and volatile) and biosurfactant concentrations were measured. The SS-SBR encouraged the growth of biosurfactant-producing species relative to the CSTR. Counts of biosurfactant-producing species (C. tropicalis, P. aeruginosa, P. fluorescens) relative to total microbial counts were 88% in the SS-SBR and 23% in the CSTR. Biosurfactants were produced in the SS-SBR to levels of nearly 70 times the critical micelle concentration (CMC) early in the cycle, but were completely degraded by the end of each cycle. No biosurfactant production was observed in the CSTR. DF biodegradation rates were over 40% greater and DF stripping was over five times lower in the SS-SBR than the CSTR. However, considerable foaming occurred in the SS-SBR. Reversing the mode of operation in the reactors on day 80 caused a complete reversal in microbial consortia and reactor performance by day 120. These results show that bioslurry reactor operation can be manipulated to control overall reactor performance.

  3. Operation control device for a nuclear reactor fuel exchanger

    International Nuclear Information System (INIS)

    Aida, Takashi.

    1984-01-01

    Purpose: To provide a operation control device for a nuclear reactor fuel exchanger with reduced size and weight capable of optionally meeting the complicated and versatile mode of the operation scope. Constitution: The operation range of a fuel exchanger is finely divided so as to attain the state capable of discriminating between operation-allowable range and operation-inhibitive range, which are stored in a memory circuit. Upon operating the fuel exchanger, the position is detected and a divided range data corresponding to the present position is taken out from the memory circuit so as to determine whether the fuel exchanger is to be run or stopped. Use of reduced size and compact IC circuits (calculation circuit, memory circuit, data latch circuit) and input/output interface circuits or the likes contributes to the size reduction of the exchanger control system to enlarge the floor maintenance space. (Moriyama, K.)

  4. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  5. RA Research nuclear reactor, Part I - RA nuclear reactor operation, maintenance and utilization in 1983

    International Nuclear Information System (INIS)

    Sotic, O.; Martinc, R.; Kozomara-Maic, S.; Cupac, S.; Raickovic, N.; Radivojevic, J.; Badrljica, R.; Majstorovic, D.; Sanovic, V.

    1983-12-01

    After regular shutdown in November 1982, inspection of the fuel elements from the RA reactor core which was done from December 1982 - February 1983 has shown that there are deposits of aluminium oxides on the surface of the fuel cladding. After restart The RA reactor was operated at power levels from 1.8 - 2 MW, with 80% enriched uranium dioxide fuel elements. It was found that there was no corrosion of the fuel element cladding and that it was not possible to find the cause of surface deposition on the cladding surfaces without further operation. It was decided to purify the heavy water permanently during operation and to increase the heavy water flow by operating two pumps. This procedure was adopted in order to decrease the possibility of corrosion. The Safety committee of the Institute has approved this procedure for operating the RA reactor in 1983. The core was made of 80% enriched fuel, critical experiments were done until June 1983, and after that the operation was continued at power levels up to 2 MW [sr

  6. Constrained model predictive control for load-following operation of APR reactors

    International Nuclear Information System (INIS)

    Kim, Jae Hwan; Lee, Sim Won; Kim, Ju Hyun; Na, Man Gyun; Yu, Keuk Jong; Kim, Han Gon

    2012-01-01

    The load-following operation of APR+ reactor is needed to control the power effectively using the control rods and to restrain the reactivity control from using the boric acid for flexibility of plant operation. Usually, the reason why the disproportion of axial flux distribution occurs during load-following operation is xenon-induced oscillation. The xenon has a very high absorption cross-section and makes the impact on the reactor delayed by the iodine precursor. The power maneuvering using automatically load-following operation has advantage in terms of safety and economic operation of the reactor, so the controller has to be designed efficiently. Therefore, an advanced control method that meets the conditions such as automatic control, flexibility, safety, and convenience is necessary to load-following operation of APR+ reactor. In this paper, the constrained model predictive control (MPC) method is applied to design APR reactor's automatic load-following controller for the integrated thermal power level and axial shape index (ASI) control. Some controllers use only the current tracking command, but MPC considers future commands in addition to the current tracking command. So, MPC can achieve better tracking performance than others. Furthermore, an MPC is to used in many industrial process control systems. The basic concept of the MPC is to solve an optimization problem for a finite future time interval at present time and to implement the first optimal control input as the current control input. The KISPAC-1D code, which models the APR+ nuclear power plants, is interfaced to the proposed controller to verify the tracking performance of the reactor power level and ASI. It is known that the proposed controller exhibits very fast tracking responses

  7. Validation of BN Reactor Plant Long-Term Operation

    International Nuclear Information System (INIS)

    Vilensky, O.; Vasilyev, B.; Kaidalov, V.

    2013-01-01

    The BN RP operation life time is mainly determined by resource of non-replaceable equipment. The new standard (RD) “Procedure of strength analysis for main components of sodium cooled fast neutron reactor plants” was developed to validate structure strength in view of radiation effects and degradation of material properties within the time period up to 300000 hours and under irradiation, as well as development of postulated crack-like defects. Using this RD, the extension of operation life of BN-600 reactor non-replaceable components from 30 to 45 years, as well as strength and durability of the most loaded non-replaceable components of BN-800 RP under construction were validated for the specified 45-year operation life. Wider application of steel 16Cr-11Ni-3Mo refers to new decisions in BN-1200 RP design that allow increasing of operation life of the most loaded non-replaceable components up to 60 years. High-chromium steel 12Cr-Ni-Mo-V-Nb is a new material, which was proposed for SG design to increase the operation life up to 30 years. In addition, the austenitic steels 18Cr-9Ni and 16Cr-11Ni-3Mo are now under upgrading for future application of them in commercial BN-1200 RP. To provide additional long-term reliable and safe operation of BN-1200 RP equipment and pipelines, it is planned to develop and implement the lifetime operational monitoring system

  8. Results of environmental radiation monitoring and meteorology measurements (material prepared for obtaining the licence for RA reactor experimental operation)

    International Nuclear Information System (INIS)

    1980-10-01

    According to the demands for obtaining the licence for restarting the Ra reactor and the experimental operation this document includes the radiation monitoring measured data in the working space and environment of the RA reactor, i.e. Boris Kidric Institute. The meteorology measured data are included as well. All the measurements are performed according to the radiation protection program applied actually from the first reactor start-up at the end of 1959 [sr

  9. Performance of a UASB reactor treating coffee wet wastewater

    International Nuclear Information System (INIS)

    Guardia Puebla, Yans; Rodríguez Pérez, Suyén; Janet Jiménez Hernández; Sánchez Girón, Víctor

    2014-01-01

    The present work shows the results obtained in the anaerobic digestion process of coffee wet wastewater processing. An UASB anaerobic reactor was operated in single-stage in mesophilic temperature controlled conditions (37±1ºC). The effect of both organic loading rate (OLR) and hydraulic retention time (HRT) in the anaerobic digestion of coffee wet wastewater was investigated. The OLR values considered in the single-stage UASB reactor varied in a range of 3,6-4,1 kgCOD m-3 d-1 and the HRT stayed in a range of 21,5-15,5 hours. The evaluation results show that the best performance of UASB reactor in single-stage was obtained at OLR of 3,6 kg COD m-3 d-1 with an average value of total and soluble COD removal of 77,2% and 83,4%, respectively, and average methane concentration in biogas of 61%. The present study suggests that the anaerobic digestion is suitable to treating coffee wet wastewater. (author)

  10. Operator performance in non-destructive testing: A study of operator performance in a performance test

    Energy Technology Data Exchange (ETDEWEB)

    Enkvist, J.; Edland, A.; Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology

    2000-05-15

    In the process industries there is a need of inspecting the integrity of critical components without disrupting the process. Such in-service inspections are typically performed with non-destructive testing (NDT). In NDT the task of the operator is to (based on diagnostic information) decide if the component can remain in service or not. The present study looks at the performance in NDT. The aim is to improve performance, in the long run, by exploring the operators' decision strategies and other underlying factors and to this way find out what makes some operators more successful than others. Sixteen operators performed manual ultrasonic inspections of four test pieces with the aim to detect (implanted) cracks. In addition to these performance demonstration tests (PDT), the operators performed independent ability tests and filled out questionnaires. The results show that operators who trust their gut feeling more than the procedure (when the two come to different results) and that at the same time have a positive attitude towards the procedure have a higher PDT performance. These results indicate the need for operators to be motivated and confident when performing NDT. It was also found that the operators who performed better rated more decision criteria higher in the detection phase than the operators who performed worse. For characterizing it was the other way around. Also, the operators who performed better used more time, both detecting and characterizing, than the operators who performed worse.

  11. Operator performance in non-destructive testing: A study of operator performance in a performance test

    International Nuclear Information System (INIS)

    Enkvist, J.; Edland, A.; Svenson, Ola

    2000-05-01

    In the process industries there is a need of inspecting the integrity of critical components without disrupting the process. Such in-service inspections are typically performed with non-destructive testing (NDT). In NDT the task of the operator is to (based on diagnostic information) decide if the component can remain in service or not. The present study looks at the performance in NDT. The aim is to improve performance, in the long run, by exploring the operators' decision strategies and other underlying factors and to this way find out what makes some operators more successful than others. Sixteen operators performed manual ultrasonic inspections of four test pieces with the aim to detect (implanted) cracks. In addition to these performance demonstration tests (PDT), the operators performed independent ability tests and filled out questionnaires. The results show that operators who trust their gut feeling more than the procedure (when the two come to different results) and that at the same time have a positive attitude towards the procedure have a higher PDT performance. These results indicate the need for operators to be motivated and confident when performing NDT. It was also found that the operators who performed better rated more decision criteria higher in the detection phase than the operators who performed worse. For characterizing it was the other way around. Also, the operators who performed better used more time, both detecting and characterizing, than the operators who performed worse

  12. Buckling resistance calculation of Guide Thimbles for the mechanical design of fuel assembly type PWR under normal reactor operating conditions

    International Nuclear Information System (INIS)

    Cruz, C.B.L.

    1990-01-01

    The calculations demonstrate the fulfillment of one of the mechanical design criteria for the Fuel Assembly Structure under normal reactor operating conditions. The calculations of stresses in the Guide Thimbles are performed with the aid of the program ANSYS. This paper contains program parameters and modelling of a typical Fuel Assembly for a Reactor similar to ANGRA II. (author)

  13. Economic evaluation of nuclear reactor operation utilizing power effect

    International Nuclear Information System (INIS)

    Budinsky, M.; Mydliar, J.

    1988-01-01

    The operation of a reactor at the so-called power effect may substantially increase the burnup of fuel to be removed. The aim of the evaluation of such reactor operation is the optimal determination of the time over which the yield of the higher use of fuel exceeds economic losses resulting from the increased share of constant expenditure of the price of generated kWh of electric power which ensues from such operation. A mathematical model is presented for such evaluation of reactor operation with regard to benefits for the national economy which is the basis of the ESTER 2 computer program. The calculations show that the prices of generated and delivered kWh are minimally 2% less than the prices of generated power without the power effect use. The minimum ranges in the interval of 30 to 50 days. The dependence of the price of generated and delivered kWh from the point of view of the operator of the power plant as well as the component of fuel price of generated kWh will not reach the minimum even after 50 days of operation. From the operating and physical points of view the duration of power effect is not expected to exceed 20 to 30 days which means that from the point of view of the national economy the price of generated and delivered kWh will be 1.6 to 2% less and the fuel component of the price of the generated kWh will be 3 to 4.5% lower. (Z.M.). 5 figs., 3 refs

  14. Research on Integration of NPP Operational Safety Management Performance Systems

    International Nuclear Information System (INIS)

    Chi, Miao; Shi, Liping

    2014-01-01

    The operational safety management of Nuclear Power Plants demands systematic planning and integrated control. NPPs are following the well-developed safety indicator systems proposed by IAEA Operational Safety Performance Indicator Programme, NRC Reactor Oversight Process or the other institutions. Integration of the systems is proposed to benefiting from the advantages of both systems and avoiding improper application into the real world. The authors analyzed the possibility and necessity for system integration, and propose an indicator system integrating method

  15. Real time monitoring system of the operation variables of the TRIGA IPR-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    Ricardo, Carla Pereira; Mesquita, Amir Zacarias

    2007-01-01

    During the last two years all the operation parameters of the TRIGA IPR-R1 were monitored and real time indicated bu the data acquisition system developed for the reactor. All the information were stored on a rigid disk, at the collection system computer, leaving the information on the reactor performance and behaviour available for consultation in a chronological order. The data acquisition program has been updated and new reactor operation parameters were included for increasing the investigation and experiments possibilities. The register of reactor operation variables are important for the immediate or subsequent safety analyses for reporting the reactor operations to the external organizations. This data acquisition satisfy the IAEA recommendations. (author)

  16. AMNT 2014. Key Topic: Reactor operation, safety - report. Pt. 1

    Energy Technology Data Exchange (ETDEWEB)

    Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany). Forschungszentrum

    2014-10-15

    Summary report on one session of the Annual Conference on Nuclear Technology held in Frankfurt, 6 to 8 May 2014: - Safety of Nuclear Installations - Methods, Analysis, Results: Backfittings for the Improvement of Safety and Efficiency. The other Sessions of the Key Topics 'Reactor Operation, Safety', 'Competence, Innovation, Regulation' and 'Fuel, Decommissioning and Disposal' will be covered in further issues of atw.

  17. Increased performance of continuous stirred tank reactor with calcium supplementation

    Energy Technology Data Exchange (ETDEWEB)

    Yuan, Zhuliang; Yang, Haijun; Zhi, Xiaohua; Shen, Jianquan [Beijing National Laboratory for Molecular Sciences (BNLMS), New Materials Laboratory, Institute of Chemistry, Chinese Academy of Sciences, Beijing 100190 (China)

    2010-04-15

    Continuous biohydrogen production with calcium supplementation at low hydraulic retention time (HRT) in a continuous stirred tank reactor (CSTR) was studied to maximize the hydrogen productivity of anaerobic mixed cultures. After stable operations at HRT of 8-4 h, the bioreactor became unstable when the HRT was lowered to 2 h. Supplementation of 100 mg/L calcium at HRT 2 h improved the operation stability through enhancement of cell retention with almost two-fold increase in cell density than that without calcium addition. Hydrogen production rate and hydrogen yield reached 24.5 L/d/L and 3.74 mol H{sub 2}/mol sucrose, respectively, both of which were the highest values our group have ever achieved. The results showed that calcium supplementation can be an effective way to improve the performance of CSTR at low HRT. (author)

  18. Ensuring the operational safety of nuclear power plants with WWER reactors

    International Nuclear Information System (INIS)

    Shasharin, G.A.; Veretennikov, G.A.; Abagyan, A.A.; Lesnoj, S.A.

    1984-01-01

    At the start of 1983, 27 nuclear power producing units with reactor facilities of the WWER type were in operation in the Soviet Union and other countries. In 1982 the average load factor for nuclear power plants with WWER reactors was 73 per cent. There was not a single nuclear accident or even damage with any significant radiation consequences in the WWER reactors during the entire period of their operation. The most modern nuclear power plants with WWER-440 and WWER-1000 reactors meet all present-day international requirements. Safe operation of the plants is achieved by a variety of measures, the most important of which include: procedures for increasing the reliability of plant equipment and systems; ensuring exact compliance with plant operating instructions; ensuring reliable operation of plant safety systems; action directed towards maintaining the skills of plant personnel at a level adequate to ensure the taking of proper action during transient processes and accident situations. The paper discusses concrete steps for ensuring safe nuclear power plant operation along these lines. In particular, measures such as the following are described: the use of a system for collecting and processing information on equipment failures and defects; the development and introduction of methods of early defect diagnosis; the performance of complex testing of safety systems; the training of highly skilled personnel for nuclear power plants at educational combines and at teaching and training centres making use of simulators; arranging accident-prevention training and special instruction for personnel. (author)

  19. Report on operation utilization and technical development of research reactors and hot laboratory

    International Nuclear Information System (INIS)

    1982-03-01

    Activities of the Division of Research Reactor Operation in fiscal 1980 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

  20. Report on operation, utilization and technical development of research reactors and hot laboratory

    International Nuclear Information System (INIS)

    1980-03-01

    Activities of the Division of Research Reactor Operation in fiscal 1978 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

  1. Report on operation, utilization and technical development of Research Reactors and Hot Laboratory

    International Nuclear Information System (INIS)

    1984-10-01

    Activities of the Division of Research Reactor Operation in fiscal 1981 are described. The division is responsible for operation and maintenance of JRR-2, JRR-3, JRR-4 and Hot Laboratory. In the above connection, various other works are performed, including technical management of fuel and coolant, radiation control, irradiation technique, etc. In Hot Laboratory, postirradiation examinations of fuels and materials are made, and also development of examination methods. (author)

  2. Nuclear Reactor RA Safety Report, Vol. 12, Accidents during reactor operation

    International Nuclear Information System (INIS)

    1986-11-01

    This volume includes description and analysis of typical accidents occurred during operation of RA reactor in chronological order, as follows: contamination of primary coolant circuit; leakage of heavy water from the primary coolant loop; contamination of vertical experimental channel; air contamination in the reactor building and loss of circulation of the primary coolant; failures of the vacuum pump and spent fuel packaging device; rupture of the spent fuel element cladding; dethronement's of capsule for irradiation of fuel element; rupture of the vertical experimental channel and contamination of the surroundings; swelling of a fuel element; appearance of deposits on the surface of the fuel elements cladding. The last chapter describes similar accidents occurred on nuclear reactors in the world [sr

  3. Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-10-15

    At the beginning of 1963 nuclear power plants produced some 3 500 000kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4-8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

  4. Operating Experience with Power Reactors. Proceedings of the Conference on Operating Experience with Power Reactors. Vol. II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1963-10-15

    At the beginning of 1963 nuclear power plants produced some 3 500 000 kW of electrical power to different distribution grids around the world. Much significant operating experience has been gained with these power reactors, but this experience is often not collected in such a way as to make it easily available. The International Atomic Energy Agency convened a Conference on Operating Experience with Power Reactors in Vienna from 4 -8 June 1963 which was attended by 240 participants representing 27 of the Agency's Member States and six international organizations. At the Conference, 42 papers giving detailed experience with more than 20 nuclear power stations were discussed. Although similar meetings on a national or regional scale have been held earlier in various countries, this is the first arranged by the Agency on a world-wide basis. Some of the detailed material may have been given earlier but for the most part it represents new and recently acquired experience, and for the first time it has been possible to compile in one place such extensive material on the operating experience with power reactors. The Conference discussed the experience gained both generally in the context of national and international nuclear power development programmes, and more specifically in the detailed operating experience with different power reactor stations. In addition, various plant components, fuel cycles, staffing of nuclear plants and licensing of such staff were treated. It is hoped that these Proceedings will be of interest not only to nuclear , plant designers and operators who daily encounter problems similar to those discussed by the Conference, but also to those guiding the planning and implementation of power development programmes.

  5. A model to describe the performance of the UASB reactor.

    Science.gov (United States)

    Rodríguez-Gómez, Raúl; Renman, Gunno; Moreno, Luis; Liu, Longcheng

    2014-04-01

    A dynamic model to describe the performance of the Upflow Anaerobic Sludge Blanket (UASB) reactor was developed. It includes dispersion, advection, and reaction terms, as well as the resistances through which the substrate passes before its biotransformation. The UASB reactor is viewed as several continuous stirred tank reactors connected in series. The good agreement between experimental and simulated results shows that the model is able to predict the performance of the UASB reactor (i.e. substrate concentration, biomass concentration, granule size, and height of the sludge bed).

  6. Evaluation of fuel performance for fresh and aged CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Bae, Jun Ho; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Like all other industrial plants, nuclear power plants also undergo degradations, so called ageing, with their operation time. Accordingly, in the recent safety analysis for a refurbished Wolsong 1 NPP, various ageing effects were incorporated into the hydraulic models of a number of the components in the primary heat transport system for conservatism. The ageing data of thermal-hydraulic components for 11 EFPY of Wolsong 1 were derived by using NUCIRC code based on the site operation data and they were modified to the appropriate input data for CATHENA code which is a thermal hydraulic code for a postulated accident analysis. This paper deals with the ageing effect of the PHTS (primary heat transport system) of CANDU reactor on the fuel performance during the normal operation. Initial conditions for fuel performance analysis were derived from the thermal-hydraulic analysis for both fresh and aged core models. Here, fresh core means a core state just right after the refurbishment and the aged core is 11 EFPY state after the refurbishment of Wolsong 1. The fuel performance was analyzed by using ELESTRES code for both fresh and aged core state and the results were compared in order to verify the ageing effect of CANDU HTS on the fuel performance.

  7. Performance of a multipurpose research electrochemical reactor

    International Nuclear Information System (INIS)

    Henquin, E.R.; Bisang, J.M.

    2011-01-01

    Highlights: → For this reactor configuration the current distribution is uniform. → For this reactor configuration with bipolar connection the leakage current is small. → The mass-transfer conditions are closely uniform along the electrode. → The fluidodynamic behaviour can be represented by the dispersion model. → This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of ±10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  8. Understanding to requirements for educational level in qualification of reactor operators

    International Nuclear Information System (INIS)

    Zhang Chi; Yang Di; Zhou Limin

    2007-01-01

    Requirements for qualification of reactor operators in nuclear safety regulations were discussed in this paper. The new issue was described in the confirmation of education level of reactor operators. The understanding to the requirements for Educational Level in Qualification of Reactor Operators was provided according to Higher Education Law of the People's Republic of China. It was proposed to improve the confirmation of qualification of reactor operators as soon as possible. (authors)

  9. Modification of the Penn State Reactor to allow transverse and rotational core motion to increase operational versatility

    International Nuclear Information System (INIS)

    Hughes, Daniel E.

    1994-01-01

    At Penn State the Nuclear Engineering students have the opportunity to perform experiments in reactor physics, work with reactor and radiation instrumentation, and operate a nuclear reactor. These activities are done at the Penn State Breazeale Reactor (PSBR), a General Atomics Mark III TRIGA reactor. Unfortunately this activity alone can not fully support the facility. The PSBR is mandated by Penn State to provide a portion of its operating budget by selling services to users outside as well as inside Penn State. In order to increase the marketability of PSBR an upgrade program was started to increase the quality and versatility of operation. The PSBR is the longest operating university reactor in the United States. The first phase of the upgrade program began in 1992. The quality of operation was increased by replacing a 1965 vintage console with a more reliable digital control and monitoring system. The present phase of the upgrade program is to increase the versatility of operation by modifying the reactor to allow transverse and rotational core motion. Adding two more degrees of motion to the reactor core increases the capability of the facility to meet the needs of present and future users. This upgrade is being financed by a grant from the Department of Energy and matching funds from Penn State. (author)

  10. Operating experiences of reactor shutdown system at MAPS

    International Nuclear Information System (INIS)

    Kotteeswaran, T.J.; Subramani, V.A.; Hariharan, K.

    1997-01-01

    The reactors in Madras Atomic Power Station (MAPS), Kalpakkam are Pressurised Heavy Water Reactors (PHWR) similar to RAPS, Kota. The moderator heavy water is pumped into the calandria from dump tank to make the reactor critical. Later with the calandria level held constant at 92% FT, the further power changes are being done with the movement of adjuster rods. The moderator is held in calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The shutdown of the reactor is effected by dumping the moderator water to dump tank by fast equalizing of helium gas pressure. In the revised mode of operation of moderator circuit after the moderator inlet manifold failure, the dump timing was observed to be more compared to the normal value. This was investigated and observed to be due to accumulation of D 2 O in the gas space above dump valves, which was affecting the helium equalizing flow. Also some of Indicating Alarm Meters (IAM) in protective system initiating the trip signals have failed in the unsafe mode. They have been modified to avoid the recurrence of the failures. (author)

  11. Systems for aiding operators at university-owned research reactors in Japan

    International Nuclear Information System (INIS)

    Nishihara, H.; Kimura, Y.; Shibata, T.

    1984-01-01

    University-owned research reactors are operated for various purposes, and small disturbances may arise from various experimental facilities. Also not uniform are the technical levels of operators who range from supervised-students to reactor physicists. Considerable efforts are therefore devoted to the preventive maintainance. With these boundary conditions imposed, systems for aiding operators are designed at these research reactor facilities. (author)

  12. Building reactor operator sustain expert system with C language integrated production system

    International Nuclear Information System (INIS)

    Ouyang Qin; Hu Shouyin; Wang Ruipian

    2002-01-01

    The development of the reactor operator sustain expert system is introduced, the capability of building reactor operator sustain expert system is discussed with C Language Integrated Production System (Clips), and a simple antitype of expert system is illustrated. The limitation of building reactor operator sustain expert system with Clips is also discussed

  13. Safe Operation of Critical Assemblies and Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1961-05-15

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  14. Safe Operation of Critical Assemblies and Research Reactors

    International Nuclear Information System (INIS)

    1961-01-01

    This Manual is provided as a guide to the safe operation of critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be of use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information. This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation of rigid standards is both impossible and undesirable at the present time, since the topics discussed form part of a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments. The Manual is intended mainly for use in those Member States where there has been little experience in the operation of critical assemblies and research reactors. It has been compounded from the best practices which exist in Member States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

  15. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    International Nuclear Information System (INIS)

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience

  16. Non-Power Reactor Operator Licensing Examiner Standards

    International Nuclear Information System (INIS)

    1994-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR Part 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, this standard will be revised periodically to accommodate comments and reflect new information or experience

  17. Non-Power Reactor Operator Licensing Examiner Standards. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-01

    The Non-Power Reactor Operator Licensing Examiner Standards provide policy and guidance to NRC examiners and establish the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55). They are intended to assist NRC examiners and facility licensees to understand the examination process better and to provide for equitable and consistent administration of examinations to all applicants by NRC examiners. These standards are not a substitute for the operator licensing regulations and are subject to revision or other internal operator examination licensing policy changes. As appropriate, these standards will be revised periodically to accommodate comments and reflect new information or experience.

  18. Ten years's reactor operation at the Technical University Zittau - operation report

    International Nuclear Information System (INIS)

    Konschak, K.

    1990-01-01

    The Zittau Training and Research Reactor ZLFR is in use for purposes of teaching the engineers who will operate the nuclear power plants in the GDR since 10 years. Since commissioning it was started up more than 1600 times, approximately two thirds of the start-ups being utilized for purposes of teaching. A number of teaching experiments were installed that demonstrate fundamental technological processes in nuclear reactors in a manner easy to understand. The high level of nuclear safety manifests itself, among other things, in extremely low radiation exposures of the operating personal and the persons to be trained. (author)

  19. N Reactor thermal plume characterization during Pu-only mode of operation

    Energy Technology Data Exchange (ETDEWEB)

    Ecker, R.M.; Thompson, F.L.; Whelan, G.

    1983-04-01

    Pacific Northwest Laboratories (PNL) performed field and modeling studies -from March 1982 through June 1983 to characterize the thermal plume from the N Reactor heated water outfall while the N Reactor operated in the Pu-only mode. Part 1 of this report deals with the field studies conducted to characterize the N Reactor thermal plume while in the Pu-only mode of operation. It includes a description of the study area, a description of field tasks and procedures, and data collection results and discussion. Part 2 describes the computer simulation of the thermal plume under different flow conditions and the calibration of the model used. It includes a description of the computer model and the assumptions on which it is based, a presentation of the input data used in this application, and a discussion of modeling results. Because the field studies were restricted by the NPOES permit variance to the spring months when high Columbia River flows prevail the mathematical modeling of the N Reactor thermal plume while the reactor operates in the Pu-only mode is instrumental in characterizing the plume during low Columbia River flows.

  20. TARMS, an on-line boiling water reactor operation management system

    International Nuclear Information System (INIS)

    Iwamoto, T.; Sakurai, S.; Uematsu, H.; Tsuiki, M.; Makino, K.

    1984-01-01

    The TARMS (Toshiba Advanced Reactor Management System) software package was developed as an effective on-line, on-site tool for boiling water reactor core operation management. It was designed to support a complete function set to meet the requirement to the current on-line process computers. The functions can be divided into two categories. One is monitoring of the present core power distribution as well as related limiting parameters. The other is aiding site engineers or reactor operators in making the future reactor operating plan. TARMS performs these functions with a three-dimensional BWR core physics simulator LOGOS 2, which is based on modified one-group, coarse-mesh nodal diffusion theory. A method was developed to obtain highly accurate nodal powers by coupling LOGOS 2 calculations with the readings of an in-core neutron flux monitor. A sort of automated machine-learning method also was developed to minimize the errors caused by insufficiency of the physics model adopted in LOGOS 2. In addition to these fundamental calculational methods, a number of core operation planning aid packages were developed and installed in TARMS, which were designed to make the operator's inputs simple and easy. (orig.) [de