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Sample records for reactor facility annual

  1. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  2. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  3. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    1995-01-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy's Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily 41 Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period

  4. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  5. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1990: Twenty-third annual report

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1993-01-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1990. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1990 annual reports submitted by about 443 licensees indicated that approximately 214,568 individuals were monitored, 110,204 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.19 rem (cSv) and an average measurable dose of about 0.36 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,361 individuals completed their employment with one or more of the 443 covered licensees during 1990. Some 77,633 of these individuals terminated from power reactor facilities, and about 11,083 of them were considered to be transient workers who received an average dose of 0.67 rem (cSv)

  6. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1992. Twenty-fifth annual report, Volume 14

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1993-12-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv)

  7. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1993. Volume 15, Twenty-six annual report

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1995-01-01

    This report the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1993. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1993 annual reports submitted by about 360 licensees indicated that approximately 189,711 individuals were monitored, 169,872 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measured dose of about 0.31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 99,749 individuals completed their employment with one or more of the 360 covered licensees during 1993. Some 91,000 of these individuals terminated from power reactor facilities, and about 12,685 of them were considered to be transient workers who received an average dose of 0.49 rem (cSv)

  8. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1995: Twenty-eighth annual report. Volume 17

    International Nuclear Information System (INIS)

    Thomas, M.L.

    1997-01-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission's (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1995 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high-level waste currently licensed, only six categories will be considered in this report. In 1995, the annual collective dose per reactor for light water reactor licensees (LWRs) was 199 person-cSv (person-rem). This is the same value that was reported for 1994. The annual collective dose per reactor for boiling water reactors (BWRs) was 256 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 170 person-cSv (person-rem). Analyses of transient worker data indicate that 17,153 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1995, the average measurable dose calculated from reported data was 0.26 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.32 cSv (rem)

  9. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1980-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1979 are described. The work of the Division is closely related to development of multi-purpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committees on Reactor Physics and on Decomissioning of Nuclear Facilities. (author)

  10. Advanced reactor experimental facilities

    International Nuclear Information System (INIS)

    Amri, A.; Papin, J.; Uhle, J.; Vitanza, C.

    2010-01-01

    For many years, the NEA has been examining advanced reactor issues and disseminating information of use to regulators, designers and researchers on safety issues and research needed. Following the recommendation of participants at an NEA workshop, a Task Group on Advanced Reactor Experimental Facilities (TAREF) was initiated with the aim of providing an overview of facilities suitable for carrying out the safety research considered necessary for gas-cooled reactors (GCRs) and sodium fast reactors (SFRs), with other reactor systems possibly being considered in a subsequent phase. The TAREF was thus created in 2008 with the following participating countries: Canada, the Czech Republic, Finland, France, Germany, Hungary, Italy, Japan, Korea and the United States. In a second stage, India provided valuable information on its experimental facilities related to SFR safety research. The study method adopted entailed first identifying high-priority safety issues that require research and then categorizing the available facilities in terms of their ability to address the safety issues. For each of the technical areas, the task members agreed on a set of safety issues requiring research and established a ranking with regard to safety relevance (high, medium, low) and the status of knowledge based on the following scale relative to full knowledge: high (100%-75%), medium (75 - 25%) and low (25-0%). Only the issues identified as being of high safety relevance and for which the state of knowledge is low or medium were included in the discussion, as these issues would likely warrant further study. For each of the safety issues, the TAREF members identified appropriate facilities, providing relevant information such as operating conditions (in- or out-of reactor), operating range, description of the test section, type of testing, instrumentation, current status and availability, and uniqueness. Based on the information collected, the task members assessed prospects and priorities

  11. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  12. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1993-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1992 (April 1, 1992-March 31, 1993). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  13. Reactor engineering department annual report

    International Nuclear Information System (INIS)

    1990-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1989 (April 1, 1989 - March 31, 1990). One of major Department's programs is the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Development of a high energy proton linear accelerator for the nuclear engineering including is also TRU incineration promoted. Other major tasks of the Department are various basic researches on nuclear data and group constants, theoretical methods and code development, on reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  14. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

    International Nuclear Information System (INIS)

    Thomas, M.L.

    1998-02-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission's (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem) 2 which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem)

  15. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-02-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem){sup 2} which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem).

  16. STARFIRE remote maintenance and reactor facility concept

    International Nuclear Information System (INIS)

    Graumann, D.W.; Field, R.E.; Lutz, G.R.; Trachsel, C.A.

    1981-01-01

    A total remote maintenance facility has been designed for all equipment located within the reactor building and hot cell, although operational flexibility has been provided by design of the reactor shielding such that personnel access into the reactor building within 24 hours after reactor shutdown is possible. The reactor design permits removal and replacement of all components if necessary, however, the vacuum pumps, isolation valves and blanket require scheduled, routine maintenance. Reactor scheduled maintenance does not dominate annual plant downtime, therefore, several scheduled operations can be added without affecting reactor availability. The maintenance facilities consist of the reactor building, the hot cell, the reactor service area and the remote maintenance control room. The reactor building contains the reactor, selected support system modules, and required maintenance equipment. The reactor and the support systems are maintained with (1) equipment that is mounted on a monorail system; (2) overhead cranes; and (3) bridge-mounted electromechanical manipulators. The hot cell is located outside of the reactor building to localize contamination products and permit independent operation. An equipment air lock connects the reactor building to the hot cell

  17. Nuclear reactor facility

    International Nuclear Information System (INIS)

    Wampole, N.C.

    1978-01-01

    In order to improve the performance of manitenance and inspections it is proposed for a nuclear reactor facility with a primary circuit containing liquid metal to provide a thermally insulated chamber, within which are placed a number of components of the primary circuit, as e.g. valves, recirculation pump, heat exchangers. The isolated placement permit controlled preheating on one hand, but prevents undesirable heating of adjacent load-bearing elements on the other. The chamber is provided with heating devices and, on the outside, with cooling devices; it is of advantage to fill it with an inert gas. (UWI) 891 HP [de

  18. Nursing Facility Initiative Annual Report

    Data.gov (United States)

    U.S. Department of Health & Human Services — This annual report summarizes impacts from the Initiative to Reduce Avoidable Hospitalizations among Nursing Facility Residents in 2014. This initiative is designed...

  19. ATR National Scientific User Facility 2013 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Ulrich, Julie A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robertson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    This is the 2013 Annual Report for the Advanced Test Reactor National Scientific User Facility. This report includes information on university-run research projects along with a description of the program and the capabilities offered researchers.

  20. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2012. Operation, utilization and technical development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

    International Nuclear Information System (INIS)

    Murayama, Yoji; Ishii, Tetsuro; Nakamura, Kiyoshi; Uno, Yuki; Ishikuro, Yasuhiro; Kawashima, Kazuhito; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Odauchi, Shouji; Maruo, Takeshi

    2014-03-01

    The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes a summary of activities of services and technical developments carried out in the period between April 1, 2012 and March 31, 2013. The activities were categorized into five service/development fields: (1) Operation and maintenance of research reactors and tandem accelerator, (2) Utilization of research reactors and tandem accelerator, (3) Upgrading of utilization techniques of research reactors and tandem accelerator, (4) Safety administration for department of research reactor and tandem accelerator, (5) International cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, number of staff members dispatched to Fukushima for the technical assistance, outcomes in service and technical developments and so on. (author)

  1. Reactor feedwater facility

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

    1996-04-30

    In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

  2. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2014. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

    International Nuclear Information System (INIS)

    Osa, Akihiko; Imahashi, Masaki; Hirane, Nobuhiko; Motome, Yuiko; Tayama, Hidekazu; Tamura, Itaru; Harada, Yuko; Sakata, Mami; Kadokura, Masakazu; Takita, Chiharu

    2017-02-01

    The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3 (Japan Research Reactor No.3), JRR-4 (Japan Research Reactor No.4), NSRR (Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2014. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration, and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

  3. Annual report of Department of Research Reactor and Tandem Accelerator, JFY2013. Operation, Utilization and Technical Development of JRR-3, JRR-4, NSRR, Tandem Accelerator and RI Production Facility

    International Nuclear Information System (INIS)

    Kashima, Yoichi; Murayama, Yoji; Nakamura, Kiyoshi; Uno, Yuki; Hirane, Nobuhiko; Ohuchi, Hitoshi; Ishizaki, Nobuhiro; Matsumura, Taichi; Nagahori, Kazuhisa; Harada, Yuko; Kadokura, Masakazu; Machi, Sumire; Takita, Chiharu

    2015-02-01

    The Department of Research Reactor and Tandem Accelerator is in charge of the operation, utilization and technical development of JRR-3(Japan Research Reactor No.3), JRR-4(Japan Research Reactor No.4), NSRR(Nuclear Safety Research Reactor), Tandem Accelerator and RI Production Facility. This annual report describes the activities of our department in fiscal year of 2013. We carried out the operation and maintenance, utilization, upgrading of utilization techniques, safety administration and international cooperation. Also contained are lists of publications, meetings, granted permissions on laws and regulations concerning atomic energy, outcomes in service and technical developments and so on. (author)

  4. Meltdown reactor core cooling facility

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1992-01-01

    The meltdown reactor core cooling facility comprises a meltdown reactor core cooling tank, a cooling water storage tank situates at a position higher than the meltdown reactor core cooling tank, an upper pipeline connecting the upper portions of the both of the tanks and a lower pipeline connecting the lower portions of them. Upon occurrence of reactor core meltdown, a high temperature meltdown reactor core is dropped on the cooling tank to partially melt the tank and form a hole, from which cooling water is flown out. Since the water source of the cooling water is the cooling water storage tank, a great amount of cooling water is further dropped and supplied and the reactor core is submerged and cooled by natural convection for a long period of time. Further, when the lump of the meltdown reactor core is small and the perforated hole of the meltdown reactor cooling tank is small, cooling water is boiled by the high temperature lump intruding into the meltdown reactor core cooling tank and blown out from the upper pipeline to the cooling water storage tank to supply cooling water from the lower pipeline to the meltdown reactor core cooling tank. Since it is constituted only with simple static facilities, the facility can be simplified to attain improvement of reliability. (N.H.)

  5. Reactor containing facility

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1992-01-01

    A cooling space having a predetermined capacity is formed between a reactor container and concrete walls. A circulation loop disposed to the outside of the concrete walls is connected to the top and the bottom of the cooling space. The circulation loop has a circulation pump and a heat exchanger, and a cooling water supply pipe is connected to the upstream of the circulation pump for introducing cooling water from the outside. Upon occurrence of loss of coolant accident, cooling water is introduced from the cooling water supply pipe to the cooling space between the reactor container and the concrete walls after shut-down of the reactor operation. Then, cooling water is circulated while being cooled by the heat exchanger, to cool the reactor container by cooling water flown in the cooling space. This can cool the reactor container in a short period of time upon occurrence of the loss of coolant accident. Accordingly, a repairing operation for a ruptured portion can be conducted rapidly. (I.N.)

  6. Reactor container facility

    International Nuclear Information System (INIS)

    Saito, Takashi; Nagasaka, Hideo.

    1990-01-01

    A dry-well pool for spontaneously circulating stored pool water and a suppression pool for flooding a pressure vessel by feeding water, when required, to a flooding gap by means of spontaneous falling upto the flooding position, thereby flooding the pressure vessel are contained at the inside of a reactor container. Thus, when loss of coolant accidents such as caused by main pipe rupture accidents should happen, pool water in the suppression pool is supplied to the flooding gap by spontaneously falling. Further, if the flooding water uprises exceeding a predetermined level, the flooding gap is in communication with the dry-well pool at the upper and the lower portions respectively. Accordingly, flooding water at high temperature heated by the after-heat of the reactor core is returned again into the flooding gap to cool the reactor core repeatedly. (T.M.)

  7. Annual report 1986 Interfacultary Reactor Institute

    International Nuclear Information System (INIS)

    1987-01-01

    In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of Current research during 198? of the departments radiochemistry, radiation chemistry, radiation physics, and reactor physics. (H.W.). refs.; figs.; tabs

  8. Reactor water injection facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro

    1997-05-02

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

  9. Emergency reactor core cooling facility

    International Nuclear Information System (INIS)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka.

    1996-01-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  10. Emergency reactor core cooling facility

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Kazuhiro; Kinoshita, Shoichiro; Iwata, Yasutaka

    1996-11-01

    The present invention provides an emergency reactor core cooling device for a BWR type nuclear power plant. Namely, D/S pit (gas/water separator storage pool) water is used as a water source for the emergency reactor core cooling facility upon occurrence of loss of coolant accidents (LOCA) by introducing the D/S pit water to the emergency reactor core cooling (ECCS) pump. As a result, the function as the ECCS facility can be eliminated from the function of the condensate storage tank which has been used as the ECCS facility. If the function is unnecessary, the level of quality control and that of earthquake resistance of the condensate storage tank can be lowered to a level of ordinary facilities to provide an effect of reducing the cost. On the other hand, since the D/S pit as the alternative water source is usually a facility at high quality control level and earthquake resistant level, there is no problem. The quality of the water in the D/S pit can be maintained constant by elevating pressure of the D/S pit water by a suppression pool cleanup (SPCU) pump to pass it through a filtration desalter thereby purifying the D/S pit water during the plant operation. (I.S.)

  11. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1985-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1984 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, safeguards technology, and activities of the Committee on Reactor Physics. (author)

  12. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Matsuura, Shojiro; Nakahara, Yasuaki; Takano, Hideki

    1982-09-01

    Research and development activities in the Division of Reactor Engineering in fiscal 1981 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  13. Reactor fuel exchanging facility

    International Nuclear Information System (INIS)

    Kubota, Shin-ichi.

    1981-01-01

    Purpose: To enable operation of an emergency manual operating mechanism for a fuel exchanger with all operatorless trucks and remote operation of a manipulator even if the exchanger fails during the fuel exchanging operation. Constitution: When a fuel exchanging system fails while connected to a pressure tube of a nuclear reactor during a fuel exchanging operation, a stand-by self-travelling truck automatically runs along a guide line to the position corresponding to the stopping position at that time of the fuel exchanger based on a command from a central control chamber. At this time the truck is switched to manual operation, and approaches the exchanger while being monitored through a television camera and then stops. Then, a manipurator is connected to the emergency manual operating mechanism of the exchanger, and is operated through necessary emergency steps by driving the snout, the magazine, the grab or the like in the exchanger in response to the problem, and necessary operations for the emergency treatment are thus performed. (Sekiya, K.)

  14. Emergency reactor container cooling facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Matsumoto, Tomoyuki.

    1992-01-01

    The present invention concerns an emergency cooling facility for a nuclear reactor container having a pressure suppression chamber, in which water in the suppression chamber is effectively used for cooling the reactor container. That is, the lower portion of a water pool in the pressure suppression chamber and the inside of the reactor container are connected by a pipeline. The lower end of the pipeline and a pressurized incombustible gas tank disposed to the outside of the reactor container are connected by a pipeline by way of valves. Then, when the temperature of the lower end of the pressure vessel exceeds a predetermined value, the valves are opened. If the valves are opened, the incombustible gas flows into the lower end of the pipeline connecting the lower portion of the water pool in the pressure suppression chamber and the inside of the reactor container. Since the inside of the pipeline is a two phase flow comprising a mixture of a gas phase and a liquid phase, the average density is decreased. Therefore, the water level of the two phase flow is risen by the level difference between the inside and the outside of the pipeline and, finally, the two phase mixture is released into the reactor container. As a result, the reactor container can be cooled by water in the suppression chamber by a static means without requiring pumps. (I.S.)

  15. Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    Schabert, H.P.; Ropers, J.

    1976-01-01

    A pressurized-water reactor pressure vessel connects via a main coolant pipe loop including a main coolant pump, with the lower portion of at least one vertical steam generator horizontally offset from the pressure vessel. This equipment is contained by a concrete structure entirely enclosing the pressure vessel and forming a generator room horizontally enclosing the generator and the loop and extending upwardly to an open top closed by a horizontal ceiling. The concrete structure is completely surrounded by a spherical steel containment shell designed to withstand any internal fluid pressure which might result from an accidental release of the coolant inside of this shell, and the shell forms a large space above the entire concrete structure. The ceiling above the generator room is a horizontal steel gridlike construction defining a plurality of vertical openings which are normally closed by horizontal sheet metal plates which are hinged to the gridlike construction and are light enough in weight to be forced upwardly, to open the openings, when the plates receive upward force from fluid pressure below them resulting from the loop, or other equipment in the generator room, accidentally permitting a sudden release of the pressurized-water coolant. The high fluid pressure that would otherwise develop within the concrete generator room, is in this way almost immediately relieved via the openings of the grid-like construction, by the plates being forced upwardly, the pressure being then dissipated upwardly in the large space above the top of the concrete structure, provided by the steel containment shell. This prevents the upstanding wall portions of the generator room from being stressed, and possibly damaged, by any sudden release of coolant in the generator room. Other features are disclosed

  16. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    Hirota, Jitsuya; Asaoka, Takumi; Suzuki, Tomoo; Mitani, Hiroshi; Akino, Fujiyoshi

    1977-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1976 are described. Works of the division concern mainly the development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and the development of Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and activities of the Committee on Reactor Physics. (auth.)

  17. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1978-10-01

    Research activities in the Division of Reactor Engineering in fiscal 1977 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committee on Reactor Physics. (Author)

  18. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1976-09-01

    Research activities conducted in Reactor Engineering Division in fiscal 1975 are summarized in this report. Works in the division are closely related to the development of multi-purpose High-temperature Gas Cooled Reactor, the development of Liquid Metal Fast Breeder Reactor by Power Reactor and Nuclear Fuel Development Corporation, and engineering research of thermonuclear fusion reactor. Many achievements are described concerning nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of the Committee on Reactor Physics. (auth.)

  19. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    1984-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1983 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  20. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1975-11-01

    Research activities in fiscal 1974 in Reactor Engineering Division of eight laboratories and computing center are described. Works in the division are closely related with the development of a multi-purpose High-temperature Gas Cooled Reactor, the development of a Liquid Metal Fast Breeder Reactor in Power Reactor and Nuclear Fuel Development Corporation, and engineering of thermonuclear fusion reactors. They cover nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and aspects of the computing center. (auth.)

  1. Pennsylvania State University Breazeale Nuclear Reactor. Thirtieth annual progress report, July 1, 1984-June 30, 1985

    International Nuclear Information System (INIS)

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

    1985-08-01

    This report is the thirtieth annual progress report of the Pennsylvania State University Breazeale Nuclear Reactor and covers such topics as: personnel; reactor facility; cobalt-60 facility; education and training; Radionuclear Application Laboratory; Low Level Radiation Monitoring Laboratory; and facility research utilization

  2. Reactor Engineering Department annual report

    International Nuclear Information System (INIS)

    Matsuura, S.; Nakahara, Y.; Takano, H.

    1983-09-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1982 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and Fusion Reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Since fiscal 1982, Systematic research and development work on safeguards technology has been added to the activities of the Department. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and safeguards technology, and activities of the Committee on Reactor Physics. (author)

  3. Annual report on JEN-1 reactor

    International Nuclear Information System (INIS)

    Montes, J.

    1972-01-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  4. Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1975-02-01

    This report summarizes main research achievements in the 48th fiscal year which were made by Reactor Engineering Division consisted of eight laboratories and Computing Center. The major research and development projects, with which the research programmes in the Division are associated, are development of High Temperature Gas Cooled Reactor for multi-purpose use, development of Liquid Metal Fast Breeder Reactor conducted by Power Reactor and Nuclear Fuel Development Corporation, and Engineering Research Programme for Thermonuclear Fusion Reactor. Many achievements are reported in various research items such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, heat transfer and fluid dynamics, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology and activities of Computing Center. (auth.)

  5. Annual report 1984 Interfacultary Reactor Institute

    International Nuclear Information System (INIS)

    1985-01-01

    In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of current research during 1984 of the departments radiochemistry, radiation chemistry, radiation physics, reactor physics, and radiation hygiene and dosimetry. (H.W.). refs.; figs.; tabs

  6. Reactor Sharing at Rensselaer Critical Facility

    International Nuclear Information System (INIS)

    D. Steiner, D. Harris, T. Trumbull

    2006-01-01

    This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

  7. Research nuclear reactor RA - Annual Report 1989

    International Nuclear Information System (INIS)

    Sotic, O.

    1989-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  8. RA Research nuclear reactor - Annual report 1987

    International Nuclear Information System (INIS)

    1987-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  9. RA Research reactor, Annual report 1988

    International Nuclear Information System (INIS)

    Sotic, O.

    1988-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  10. Reactor incident status 1981 annual report

    International Nuclear Information System (INIS)

    Kiser, S.H.

    1982-01-01

    Reactor Incident followup action is summarized through periodic status reports. This annual report summarizes action taken or anticipated for Reactor Incidents through December 1981. Incidents for which action has been completed, have been deleted from the report. Quarterly addende will update the report by tabulating incidents for each three month period through the coming year. The report consists of a part for the P, K, and C Reactors. Each reactor part is divided into three sections: Further Technical Analysis or Followup Needed; Funding and/or Implementation Needed; and No Further Technical Analysis Anticipated

  11. 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

  12. Annual report 1983 Interfacultary Reactor Institute

    International Nuclear Information System (INIS)

    1984-01-01

    In this annual report of the Dutch Interfacultary Reactor Institute, summary- and detailed reports are presented of current research during 1983 of the departments radiochemistry, radiation chemistry, radiation physics, reacotor physics, and radiation hygiene and dosimetry. (H.W.). refs.; figs.; tabs

  13. Research nuclear reactor RA - Annual report 1992

    International Nuclear Information System (INIS)

    Sotic, O.

    1992-12-01

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

  14. Irradiation Facilities at the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2005-01-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

  15. RB research nuclear reactor, Annual report for 1984, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.; Ilic, I.

    1984-01-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source

  16. Research nuclear reactor RA - Annual Report 1994

    International Nuclear Information System (INIS)

    Sotic, O.

    1994-12-01

    Activities related to revitalisation of the RA reactor stared in 1986, were continued in 1991. A number of interventions on the reactor components were finished that are supposed to enable continuous and reliable operation. The last, and at the same time largest action, related to exchange of complete reactor instrumentation is underway, but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Activities related to improvement of Russian project were continued in 1994. Control and maintenance of the reactor components was done regularly and efficiently. Extensive repair of the secondary coolant loop is almost finished and will be completed in the first part of 1995 according to existing legal procedures and IAEA recommendations. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 47 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. Research reactor RA Annual report for year 1991 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection [sr

  17. RA Research reactor, Annual report 1971

    International Nuclear Information System (INIS)

    Milosevic, D. et al.

    1971-12-01

    During 1971, the RA Reactor was operated at nominal power of 6.5 MW for 190 days, and 50 days at lower power levels. Total production mounted to 31606 MWh which is 5.3% higher than planned, and the highest annual level since the reactor started operation. Reactor was used for irradiation and experiments according to the demand of 425 users, of which 370 from the Institute and 55 external users. This report contains detailed data about reactor power and experiments performed in 1971. Discrepancies from the action plan, meaning higher production was achieved in June and December due to special demand of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to power cuts during reactor operation. There was no longer interruption caused by failures of the equipment. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. Shorter interruptions resulted from breaking of connectors in the technical water pipe system caused by soil sliding near the pumping station on the Danube. Total personnel exposure dose was lower than during previous years. There was no accident nor any event that could be called accidental. Decontamination od surfaces was less than during previous years. It was concluded that the successful operation in 1971 resulted from efficient work during past years. But, some of the activities were interrupted due to undefined policy concerned with operation of the RA reactor and financial issues. This involves study of the possibility to use highly enriched fuel that would increase the useful neutron flux and the reactor compatibility with similar reactors for the future ten years. Another project that has been interrupted is construction of the emergency core cooling system which is important for the reactor safety. Financial problems have influenced not only the reactor operation but the number of employees, which could cause negative

  18. Research nuclear reactor RA, Annual Report 2001

    International Nuclear Information System (INIS)

    Sotic, O.

    2002-01-01

    During 2001, activities at the RA research nuclear reactor in were performed according to the Contract about financing of the RA reactor for the period January-December 2001, signed by the Ministry of Science, technology and development of the Republic of Serbia. RA reactor was not operated since shutdown in August 1984. Although, the most of the planned reconstruction activities were finished until 1991, the most important, which was concerned with exchange of the reactor instrumentation, financed by the IAEA, was interrupted due to international sanctions imposed on the country. Since 1992, all the renewal and reconstruction activities were ceased. Continuous aging and degradation of the equipment and facilities demand decision making about the future status of the Ra reactor. Until this decision is made it is an obligation to maintain control and maintenance of ventilation system, power supply, internal transportation system, spent fuel storage, hot cells, electronic fuel surveillance system, and part of the stationary dosimetry system. In 2001, apart from the mentioned activities, actions were undertaken related to maintenance of the reactor building and installations. The most important tasks fulfilled were: protection of the roof of the ventilation system building, purchase and installing the fire protection system and twelve new battery cells in the reactor building. There were no actions concerned with improvement of the conditions for intermediate spent fuel storage. With the support of IAEA, actions were initiated for possible transport of the spent fuel tu Russia. At the end of 2001, preparations were started for possible future decommissioning of the RA reactor. After, renewal of the membership of our country in the IAEA, the Government of Yugoslavia has declared its attitude about the intention of RA reactor decommissioning at the General Conference in September 2001 [sr

  19. Research nuclear reactor RA - Annual Report 1991

    International Nuclear Information System (INIS)

    Sotic, O.

    1992-01-01

    Activities related to revitalisation of the RA reactor stared in 1986, were continued in 1991. A number of interventions on the reactor components were finished that are supposed to enable continuous and reliable operation. The last, and at the same time largest action, related to exchange of complete reactor instrumentation is underway, but it is behind the schedule in 1991 because the delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. In 1991 most of the existing RA reactor instrumentation was dismantled, only the part needed for basic measurements when reactor is not operated, was maintained. Construction of some support elements is almost finished by the local staff. The Institute has undertaken this activity in order to speed up the ending of the project. If all the planned instrumentation would not arrive until the end of March 1992, it would not be possible to start the RA reactor testing operation in the first part of 1993, as previously planned. In 1991, 53 staff members took part in the activities during 1991, which is considered sufficient for maintenance and repair conditions. Research reactor RA Annual report for year 1991 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection [sr

  20. Imperial College Reactor Centre annual report. 1984

    International Nuclear Information System (INIS)

    1985-01-01

    The following new equipment is noted; for atomic absorption spectrometry to supplement the neutron activation analysis, and an additional nuclear data analysis system to improve the quality and speed of the service to users of the Centre's facilities. Users include undergraduates from the University of London, outside bodies such as the British Musueum, as well as departments of Colleges of the University of London. The reactor lost only three days through failures or faults. Two replacement fuel elements were put into the reactor during the year. The report contains brief accounts of 34 research programmes at the Centre. (U.K.)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  2. Estimated Annual Maintenance Costs for Educational Facilities in Eritrea

    DEFF Research Database (Denmark)

    Vagnby, Bo Hellisen

    Global estimated annual costs for undertaking basic maintenance of all educational facilities in Eritrea.......Global estimated annual costs for undertaking basic maintenance of all educational facilities in Eritrea....

  3. RB research nuclear reactor - Annual report for 1986, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1987-01-01

    This report includes data concerning the RB reactor operation in 1986, state of the reactor components, data about the employed personnel and the database of experimental and other reactor related devices. It is made of 3 parts: Engineering description and operation of the RB reactor including dosimetry, reactor staff data and financial report; Reactor facility components and maintenance; RB reactor operation and utilization in 1986 [sr

  4. Brennilis nuclear facilities. 2009 annual report

    International Nuclear Information System (INIS)

    2010-01-01

    This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Brennilis, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

  5. Civaux nuclear facilities. 2009 annual report

    International Nuclear Information System (INIS)

    2010-01-01

    This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Civaux, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

  6. Chooz nuclear facilities. 2009 annual report

    International Nuclear Information System (INIS)

    2010-01-01

    This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Chooz, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

  7. RB research nuclear reactor, Annual report for 1983, I - III

    International Nuclear Information System (INIS)

    Markovic, H.; Pesic, M.; Vranic, S.; Petronijevic, M.; Zivkovic, B.

    1983-01-01

    The annual report for 1981 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff; financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; utilization of the reactor as a radiation source. It contains the preliminary safety report for operating the reactor with the internal neutron converter and the plan for criticality experiment with the converter

  8. Claim prevention at reactor facilities

    International Nuclear Information System (INIS)

    Colby, B.P.

    1987-01-01

    Why does a radiation worker bring a claim alleging bodily injury from radiation exposure? Natural cancer, fear of radiation induced cancer, financial gain, emotional distress and mental anguish are some reasons for workers' claims. In this paper the author describes what power reactor health physicists are doing to reduce the likelihood of claims by establishing programs which provide sound protection of workers, prevent radiological events, improve workers' knowledge of radiological conditions and provide guidance for radiological incident response

  9. Effluent releases at the TRIGA reactor facility

    Energy Technology Data Exchange (ETDEWEB)

    Whittemore, W L [General Atomic Co., San Diego, CA (United States)

    1974-07-01

    The principal effluent from the operating TRIGA reactors in our facility is argon-41. As monitored by a recording gas and particulate stack monitor, the values shown in the table, the Mark III operating 24 hours per day for very long periods produced the largest amount of radioactive argon. The quantity of 23.7 Ci A-41 when diluted by the normal reactor room ventilation system corresponded to 1.45 x 10{sup -6} {mu}Ci/cc. As diluted in the roof stack stream and the reactor building wake, the concentration immediately outside the reactor building was 25% MPC for an unrestricted area. The continued dilution of this effluent resulted in a concentration of a few percent MPC at the site boundary (unrestricted area) 350 meters from the reactor. (author)

  10. SERC Central Laser Facility annual report 1992

    International Nuclear Information System (INIS)

    1992-01-01

    In this 1992 Annual Report to the Laser Facility Committee of the Science and Engineering Research Council, the Central Laser Facility at Rutherford Appleton Laboratory, technical progress is described and mid-term organizational goals outlined. Outstanding among recent achievements is the work on plasma heating being undertaken on the Sprite facility using the ultra-bright KrF laser pumped Raman beams. Two-beam operation at power levels approaching 2 TW in 10 ps are hoped for. On a four year timescale the Titania system will provide four Raman beams of exceptional brightness and power up to 20TW in 10ps. The other high power laser facility, Vulcan is also producing exciting work. Progress in nanosecond studies using Raman spectroscopy have produced the first Raman spectrum of solvated Buckmister fullerene and direct observation of the separation of germinate ion pairs, as well as information on the behaviour of a single base in an oligonuclide chain. Phase boundaries for the solidification of a two dimensional electron fluid have been determined in a Gallium Arsenide heterojunction. Despite staff number attrition, operation and development of the facilities have continued successfully. (UK)

  11. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  12. Advanced Test Reactor National Scientific User Facility

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

    2011-01-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  13. Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  14. Cooling facility for reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro

    1996-05-31

    A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

  15. Cooling facility for reactor container

    International Nuclear Information System (INIS)

    Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro.

    1996-01-01

    A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

  16. Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1979-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1978 are described. Works of the Division are development of multi-purpose Very High Temperature Gas Cooled Reactor, fusion reactor engineering, and development of Liquid Metal Fast Breeder Reactor for Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, dynamics analysis and control method development, fusion reactor technology, and Committees on Reactor Physics and in Decommissioning of Nuclear Facilities. (author)

  17. 1980 Annual status report reactor safety

    International Nuclear Information System (INIS)

    1981-01-01

    The JRC reactor safety programme involves theoretical and experimental activities to analyse accidents and their consequences for LWRs and LMFBRs. The first project deals with the improvement and the application of methodologies for risk and reliability assessment. This activity involves the identification and modelling of accident sequences and events and the analysis of fault trees. In this project, the implementation of a centralized data bank system (European Reliability Data System) is foreseen, which should provide the information needed for risk assessment studies. In project 2 a major effort on LWRs is centered on the study of the loss-of-coolant accident following large, intermediate or small breaks of the primary circuit. These accidents are simulated out of pile in the LOBI facility. In project 3 a contribution is made to solve material problems and to provide data and calculation methods for end of life predictions of reactor components. It involves a contribution to the programme for the inspection of steel components (PISC) as well as the study of fracture and creep fatigue properties of stainless steel. In the project 4 and 5 a deterministic approach is adopted to solve some problems of large hypothetical accidents in an LMFBR. The calculation tools developed concern sodium thermohydraulics in fuel element bundles, fuel coolant interaction, whole core accident analysis, containment loading and response and post accident heat removal

  18. Decommissioning of reactor facilities (2). Required technology

    International Nuclear Information System (INIS)

    Yanagihara, Satoshi

    2014-01-01

    Decommissioning of reactor facilities was planned to perform progressive dismantling, decontamination and radioactive waste disposal with combination of required technology in a safe and economic way. This article outlined required technology for decommissioning as follows: (1) evaluation of kinds and amounts of residual radioactivity of reactor facilities with calculation and measurement, (2) decontamination technology of metal components and concrete structures so as to reduce worker's exposure and production of radioactive wastes during dismantling, (3) dismantling technology of metal components and concrete structures such as plasma arc cutting, band saw cutting and controlled demolition with mostly remote control operation, (3) radioactive waste disposal for volume reduction and reuse, and (4) project management of decommissioning for safe and rational work to secure reduction of worker's exposure and prevent the spreading of contamination. (T. Tanaka)

  19. RB Research nuclear reactor RB reactor, Annual report for 2000

    International Nuclear Information System (INIS)

    Milosevic, M.

    2000-12-01

    Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

  20. University of Arizona TRIGA reactor. Annual utilization report, 1984-1985

    International Nuclear Information System (INIS)

    Nelson, G.W.

    1986-01-01

    This is the annual report for the University of Arizona TRIGA Reactor under Contract No. DE-AC02-76ER02096 covering the period July 1, 1984 through June 30, 1985, including the 1984-85 Academic Year. The purpose of this report is to document the facility usage which is possible because of DOE support under the contract. The reactor is operated under License R-52 with the United States Nuclear Regulatory Commission

  1. Annual report 1986 of Society for Reactor Safety (GRS)

    International Nuclear Information System (INIS)

    1987-01-01

    The overall development of Society for Reactor Safety (GRS) is indicated by its annual volume of achievement. After initially high growth rates that achievement has come to stand at some 62 million marks or so over the last years. GRS today employs some 450 staff, of which 325 are scientific-technical experts representing altogether nearly 4000 man-years of experience in nuclear technology. The report presents essential results from the main points of activity of GRS over 10 years. In accordance with the aim of the institution they primarily cover projects in the sectors safety and radiation protection for nuclear facilities. Since 1985, since the extension of its contract of associates, GRS has also increasingly been concerned with safety questions in non-nuclear technology, concentrating mainly on the transfer of existing methods. (orig./DG) [de

  2. Annual report of the CTR Blanket Engineering research facility in 1996

    International Nuclear Information System (INIS)

    1998-02-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1996. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

  3. Annual report of the CTR Blanket Engineering research facility in 1992

    International Nuclear Information System (INIS)

    1993-08-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1992. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (J.P.N.)

  4. Annual report of the CTR Blanket Engineering research facility in 1994

    International Nuclear Information System (INIS)

    1995-09-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor(CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1994. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  5. Annual report of the CTR blanket engineering research facility in 1993

    International Nuclear Information System (INIS)

    1994-08-01

    This is an annual report of the studies on Controlled Thermo-nuclear Reactor (CTR) Blanket Engineering which have been carried out in the Faculty of Engineering, the University of Tokyo, in FY 1993. This research facility on the CTR Blanket Engineering is located in the Nuclear Engineering Research Laboratory, the Tokai-mura branch of the Faculty of Engineering. (author)

  6. Emergency facility control device for nuclear reactor

    International Nuclear Information System (INIS)

    Ikehara, Morihiko.

    1981-01-01

    Purpose: To increase the reliability of a nuclear reactor by allowing an emergency facility to be manually started and stopped to make its operation more convenient and eliminate the possibility of erroneous operation in an emergency. Constitution: There are provided a first water level detector for detecting a level lower than the first low water level in a reactor container and a second water level detector for detecting a level lower than the second low water level lower than the first low water level, and an emergency facility can be started and stopped manually only when the level is higher than the second low water level, but the facility will be started regardless of the state of the manual operation when the level is lower than the second low water level. Thus, the emergency facility can be started by manual operation, but will be automatically started so as to secure the necessary minimum operation if the level becomes lower than the second low water level and the stopping operation thereafter is forgotten. (Kamimura, M.)

  7. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  8. Imperial College Reactor Centre annual report 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The report covers the following matters: research topics (reactor engineering; neutron and gamma dosimetry; nuclear physics; stable and radiotracer studies; neutron activation analysis (medicine; the environment; archaeology; geology)); personnel; publications; overseas visits; research contracts; teaching; reactor operations. (U.K.)

  9. Oregon State University TRIGA Reactor annual report

    Energy Technology Data Exchange (ETDEWEB)

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

    1979-08-31

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included.

  10. Oregon State University TRIGA Reactor annual report

    International Nuclear Information System (INIS)

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

    1979-01-01

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included

  11. Experimental facilities for Generation IV reactors research

    International Nuclear Information System (INIS)

    Krecanova, E.; Di Gabriele, F.; Berka, J.; Zychova, M.; Macak, J.; Vojacek, A.

    2013-06-01

    Centrum Vyzkumu Rez (CVR) is research and development Company situated in Czech Republic and member of the UJV group. One of its major fields is material research for Generation IV reactor concepts, especially supercritical water-cooled reactor (SCWR), very high temperature/gas-cooled fast reactor (VHTR/GFR) and lead-cooled fast reactor (LFR). The CVR is equipped by and is building unique experimental facilities which simulate the environment in the active zones of these reactor concepts and enable to pre-qualify and to select proper constructional materials for the most stressed components of the facility (cladding, vessel, piping). New infrastructure is founded within the Sustainable Energy project focused on implementation the Generation IV and fusion experimental facilities. The research of SCWR concept is divided to research and development of the constructional materials ensured by SuperCritical Water Loop (SCWL) and fuel components research on Fuel Qualification Test loop (SCWL-FQT). SCWL provides environment of the primary circuits of European SCWR, pressure 25 MPa, temperature 600 deg. C and its major purpose is to simulate behavior of the primary medium and candidate constructional materials. On-line monitoring system is included to collect the operational data relevant to experiment and its evaluation (pH, conductivity, chemical species concentration). SCWL-FQT is facility focused on the behavior of cladding material and fuel at the conditions of so-called preheater, the first pass of the medium through the fuel (in case of European SCWR concept). The conditions are 450 deg. C and 25 MPa. SCWL-FQT is unique facility enabling research of the shortened fuel rods. VHTR/GFR research covers material testing and also cleaning methods of the medium in primary circuit. The High Temperature Helium Loop (HTHL) enables exposure of materials and simulates the VHTR/GFR core environment to analyze the behavior of medium, especially in presence of organic compounds and

  12. Lining facility for FBR type reactor

    International Nuclear Information System (INIS)

    Shimano, Kunio.

    1991-01-01

    In a lining facility for protecting structural material concretes for concrete buildings in an FBR type power plant, sodium-resistant and heat-resistant first and second coating layers are lined at the surface of concretes, and steam releasing materials are disposed between the first and the second coating layers for releasing water contents evaporated from the concretes to the outside. With such a constitution, since there is no structures for welding steel plates to each other as in the prior art, the fabrication is made easy. Further, since cracks of coating materials can be suppressed, reactor safety is improved. (T.M.)

  13. RB research nuclear reactor, Annual report for 1981

    International Nuclear Information System (INIS)

    Markovic, H.; Sotic, O.; Pesic, M.; Vranic, S.; Zivkovic, B.; Bogdanovic, M.; Petronijevic, M.

    1981-01-01

    The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report

  14. Risk management activities at the DOE Class A reactor facilities

    International Nuclear Information System (INIS)

    Sharp, D.A.; Hill, D.J.; Linn, M.A.; Atkinson, S.A.; Hu, J.P.

    1993-01-01

    The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented

  15. Ground test facility for nuclear testing of space reactor subsystems

    International Nuclear Information System (INIS)

    Quapp, W.J.; Watts, K.D.

    1985-01-01

    Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

  16. Facility Effluent Monitoring Plan for the N Reactor

    International Nuclear Information System (INIS)

    Watson, D.J.; Brendel, D.F.; Shields, K.D.

    1991-11-01

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP- 0438. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan is the first annual report. It shall ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated as a minimum every three years. The primary purpose of the N Reactor Facility Effluent Monitoring Plan (FEMP), during standby, is to ensure that the radioactive effluents are properly monitored and evaluated for compliance with the applicable DOE orders and regulatory agencies at the federal, state, and local levels. A secondary purpose of the FEMP is to ensure that hazardous wastes are not released, in liquid effluents, to the environment even though the potential to do so is extremely low. The FEMP is to provide a monitoring system that collects representative samples in accordance with industry standards, performs analyses within stringent quality control (QC) requirements, and evaluates the data through the use of comparative analysis with the standards and acceptable environmental models

  17. Annual report on JEN-1 and JEN-2 Reactors

    International Nuclear Information System (INIS)

    Montes Ponce de Leon, J.

    1974-01-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  18. Nuclear Reactor Laboratory annual report, fiscal year 1981-1982

    International Nuclear Information System (INIS)

    Cashwell, R.J.

    1982-01-01

    Information related to the use of the UWNR reactor is presented concerning instructional use by the Nuclear Engineering Department; reactor sharing program; utility personnel training; sample irradiations and neutron activation analysis services; changes in personnel, facility, and procedures; and results of surveillance tests

  19. Decommissioning the UHTREX Reactor Facility at Los Alamos, New Mexico

    International Nuclear Information System (INIS)

    Salazar, M.; Elder, J.

    1992-08-01

    The Ultra-High Temperature Reactor Experiment (UHTREX) facility was constructed in the late 1960s to advance high-temperature and gas-cooled reactor technology. The 3-MW reactor was graphite moderated and helium cooled and used 93% enriched uranium as its fuel. The reactor was run for approximately one year and was shut down in February 1970. The decommissioning of the facility involved removing the reactor and its associated components. This document details planning for the decommissioning operations which included characterizing the facility, estimating the costs of decommissioning, preparing environmental documentation, establishing a system to track costs and work progress, and preplanning to correct health and safety concerns in the facility. Work to decommission the facility began in 1988 and was completed in September 1990 at a cost of $2.9 million. The facility was released to Department of Energy for other uses in its Los Alamos program

  20. ATR National Scientific User Facility 2015 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kennedy, J. Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ogden, Dan [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cole, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Knight, Collin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Teysseyre, Sebastien [Idaho National Lab. (INL), Idaho Falls, ID (United States); Benson, Jeff [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidrich, Brenden [Idaho National Lab. (INL), Idaho Falls, ID (United States); Jackson, John [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bean, Lindy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This is the 2015 Annual Report for the Nuclear Science User Facility. This report includes information on university-run research projects along with a description of the program and the capabilities offered researchers.

  1. 2015 Annual Report - Argonne Leadership Computing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Collins, James R. [Argonne National Lab. (ANL), Argonne, IL (United States); Papka, Michael E. [Argonne National Lab. (ANL), Argonne, IL (United States); Cerny, Beth A. [Argonne National Lab. (ANL), Argonne, IL (United States); Coffey, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-01

    The Argonne Leadership Computing Facility provides supercomputing capabilities to the scientific and engineering community to advance fundamental discovery and understanding in a broad range of disciplines.

  2. 2014 Annual Report - Argonne Leadership Computing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Collins, James R. [Argonne National Lab. (ANL), Argonne, IL (United States); Papka, Michael E. [Argonne National Lab. (ANL), Argonne, IL (United States); Cerny, Beth A. [Argonne National Lab. (ANL), Argonne, IL (United States); Coffey, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-01-01

    The Argonne Leadership Computing Facility provides supercomputing capabilities to the scientific and engineering community to advance fundamental discovery and understanding in a broad range of disciplines.

  3. Nuclear Safety Research and Facilities Department. Annual report 1999

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E. [eds.

    2000-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  4. Nuclear Safety Research and Facilities Department annual report 1999

    DEFF Research Database (Denmark)

    Majborn, B.; Damkjær, A.; Jensen, Per Hedemann

    2000-01-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department´s research and development activities were organized in two research programmes: "Radiation Protection and Reactor Safety" and"Radioecology and Tracer Studies". The nuclear...... facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are includedtogether with a summary of the staff´s participation in national and international committees....

  5. Nuclear Safety Research and Facilities Department annual report 1997

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Aarkrog, A.; Brodersen, K. [and others

    1998-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1997. The department`s research and development activities were organized in four research programmes: Reactor Safety, Radiation protection, Radioecology, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the educational reactor DR1. Lists of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au) 11 tabs., 39 ills.; 74 refs.

  6. Nuclear Safety Research and Facilities Department annual report 1998

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E

    1999-04-01

    The report present a summary of the work of the Nuclear Safety Research and Facilities Department in 1998. The department`s research and development activities were organized in two research programmes: `Radiation Protection and Reactor Safety` and `Radioecology and Tracer Studies`. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment plant, and the educational reactor DR1. Lsits of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au)

  7. Nuclear Safety Research and Facilities Department. Annual report 1999

    International Nuclear Information System (INIS)

    Majborn, B.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E.

    2000-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1999. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR 3, the Isotope Laboratory, the Waste Management Plant, and the educational reactor DR 1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  8. Nuclear Safety Research and Facilities department annual report 1996

    International Nuclear Information System (INIS)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Heydorn, K.; Oelgaard, P.L.

    1997-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1996. The Department's research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au) 2 tabs., 28 ills

  9. Nuclear Safety Research and Facilities Department annual report 1997

    International Nuclear Information System (INIS)

    Majborn, B.; Aarkrog, A.; Brodersen, K.

    1998-04-01

    The report presents a summary of the work of the Nuclear Safety Research and Facilities Department in 1997. The department's research and development activities were organized in four research programmes: Reactor Safety, Radiation protection, Radioecology, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the educational reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  10. Nuclear Safety Research and Facilities Department annual report 1998

    International Nuclear Information System (INIS)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Hedemann Jensen, P.; Nielsen, S.P.; Nonboel, E.

    1999-04-01

    The report present a summary of the work of the Nuclear Safety Research and Facilities Department in 1998. The department's research and development activities were organized in two research programmes: 'Radiation Protection and Reactor Safety' and 'Radioecology and Tracer Studies'. The nuclear facilities operated by the department include the research reactor DR3, the Isotope Laboratory, the Waste Treatment plant, and the educational reactor DR1. Lsits of staff and publications are included together with a summary of the staff's participation in national and international committees. (au)

  11. 1982 annual status report: reactor safety

    International Nuclear Information System (INIS)

    1982-01-01

    This report presents the projects of the Reactor Safety Program at the JRC: 1) Reliability and risk evolution; 2) LWR loss of coolant accident studies; 3) Primary system integrity; 4) LMFBR core accident initiation and transition phase; and, 5) LMFBR accident post disassembly phase

  12. RA Research reactor, Annual report 1969

    International Nuclear Information System (INIS)

    Milosevic, D. et al.

    1969-12-01

    During 1969, the RA Reactor was operated at nominal power of 6.5 MW for 200 days, and 15 days at lower power levels. Total production mounted to 31131 MWh which is 3.77% higher than planned. Reactor was used for irradiation and experiments according to the demand of 463 users from the Institute and 63 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. If there had been no problems with power supply during last three months and Danube low water level in September and October the past year would have been the most successful up to now. The number od scram shutdowns was not higher than during past two years in spite of the difficulties in the last quarter. There were three incidents which caused higher personnel exposure during operation. One, was the destruction of the canner with silver (because the time spent in the core was too long) which caused the surface contamination of the platform, the background radiation was 10 to 100 times higher than regular. The other two cases were caused by failure of the device for handling the fuel slugs in the fuel channels during refuelling. Reactor refuelling was done four times during 1969, and 499 fresh fuel slugs were used. Refuelling applied the approach of 'mixing' the fresh fuel slugs with the 'old' fuel slugs in the fuel channel. Decontamination of surfaces was on the same level as previously in spite of the problems with silver. Since two staff members have left, the present number od employees is now the minimum needed for reactor operation and maintenance. It is stated that the operation of components and equipment is on sufficiently high level after ten years of reactor operation. The action plan for 1970 is made according to the same principles as in previous four years but the planned production is decreased to 25000 MWh, because control of important components is needed after ten

  13. Annual report on JEN-1 reactor; Informe periodico del Reactor JEN-1 correspondiente al ano 1971

    Energy Technology Data Exchange (ETDEWEB)

    Montes, J

    1972-07-01

    In the annual report on the JEN-1 reactor the main features of the reactor operations and maintenance are described. The reactor has been critical for 1831 hours, what means 65,8% of the total working time. Maintenance and pool water contamination have occupied the rest of the time. The maintenance schedule is shown in detail according to three subjects. The main failures and reactor scrams are also described. The daily maximum values of the water activity are given so as the activity of the air in the reactor hall. (Author)

  14. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  15. Present status of decommissioning in the Musashi Reactor Facility (4)

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Tanzawa, Tomio; Mitsuhashi, Ishi; Morishima, Kayoko; Matsumoto, Tetsuo

    2012-01-01

    The decommissioning of the Musashi reactor was decided in 2003. Permanent shutdown of the reactor and stopping the operational functions were conducted in 2004. Transportation of the spent fuels was finished in 2006. After 2007, the system and equipment stopping the functions were stored as installed in the reactor facility as radioactive wastes. After separating nonradioactive wastes such as concretes from radioactive wastes with a contamination test, stopping the functions of liquid waste management facility was performed with newly installed drainage facility for radioisotope use in 2010. Solid waste management facility was also dismantled and removed in the same way as liquid waste management facility in 2011. Radioactive wastes packed in containers were moved and stored in the reactor facility. (T. Tanaka)

  16. Imperial College Reactor Centre annual report. 1983

    International Nuclear Information System (INIS)

    1984-01-01

    It is reported that the reactor operated reliably during the year with less than half a day of operating time lost by faults or failures. Brief accounts of the 34 research projects at the Centre are given, and a list of teaching experiments or visits is included. These include undergraduate and post-graduate teaching. Commercial requests for irradiations and neutron activation analysis are reported as increasing. (U.K.)

  17. RA Research reactor, Annual report 1972

    International Nuclear Information System (INIS)

    Milosevic, D. et al.

    1972-12-01

    During 1972, the total production was 31151 MWh which is 3.8% higher than planned. The reactor was used for irradiation and experiments according to the demand of 381 users, of which 340 from the Institute and 41 external users. This report contains detailed data about reactor power and experiments performed in 1972. Discrepancies from the action plan, meaning higher production was achieved due to special demands of the users. Total number of interruptions was lower than during all the previous years, and were caused mainly due to announced power cuts. There was only on scram shutdown during this year caused by a false signal of the reactor control instrumentation. There were no longer interruptions. One shorter interruption (shorter than 24 hours) caused by removal of a UO 2 capsule from the core, placed there for measuring heat transfer. Total personnel exposure dose was lower than during previous years. One accident caused contamination with gases and aerosols containing mainly shot-living isotopes. Decontamination od surfaces was less than during previous years. Practically there was no surface contamination that would demand action of the decontamination team, except for the regular decontamination after refueling. It was concluded that the successful operation in 1972 has a special significance having taking in account the financial crisis caused by the unresolved status of the reactor. It is emphasised, in the plan for the next year that there is an urgent need of making a long-term plan of rector application. It is indispensable to finish preparatory tasks for replacing the fuel with the highly enriched fuel elements by 1974, and building the core emergency cooling system [sr

  18. Power Reactor Docket Information. Annual cumulation (citations)

    International Nuclear Information System (INIS)

    1977-12-01

    An annual cumulation of the citations to the documentation associated with civilian nuclear power plants is presented. This material is that which is submitted to the U.S. Nuclear Regulatory Commission in support of applications for construction and operating licenses. Citations are listed by Docket number in accession number sequence. The Table of Contents is arranged both by Docket number and by nuclear power plant name

  19. 2016 Annual Report - Argonne Leadership Computing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Collins, Jim [Argonne National Lab. (ANL), Argonne, IL (United States); Papka, Michael E. [Argonne National Lab. (ANL), Argonne, IL (United States); Cerny, Beth A. [Argonne National Lab. (ANL), Argonne, IL (United States); Coffey, Richard M. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-01-01

    The Argonne Leadership Computing Facility (ALCF) helps researchers solve some of the world’s largest and most complex problems, while also advancing the nation’s efforts to develop future exascale computing systems. This report presents some of the ALCF’s notable achievements in key strategic areas over the past year.

  20. Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

    International Nuclear Information System (INIS)

    1995-01-01

    With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor's Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced

  1. Study of fast reactor safety test facilities. Preliminary report

    International Nuclear Information System (INIS)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods

  2. Magnox Electric Littlebrook reactor inspection and repair rehearsal facility

    International Nuclear Information System (INIS)

    Barnes, S.A.; Clayton, R.; Gaydon, B.G.; Ramsey, B.H.

    1996-01-01

    Magnox reactors, although designed to be maintenance free during their operational life, have nevertheless highlighted the need for test rig facilities to train operators in the methods and techniques of reactor inspection and repair. The history of the facility for reactor engineering development (FRED) is described and its present role as a repair rehearsal facility noted. Advances in computer graphics may, in future, mean that such operator training will be virtual reality rather than analog reality based; however the need for such rigs to commission techniques and equipment and to establish performance and reliability is likely to continue. (UK)

  3. OSU Reactor Sharing Program FY 1995 annual report

    International Nuclear Information System (INIS)

    Higginbotham, J.F.

    1996-10-01

    This is the annual report of the activities supported under the Oregon State University Reactor Sharing Program, award number DE-FG06-NE38137. The beginning date for the award was September, 30, 1995 and the end date was September 29, 1996. Work conducted under this award is internally administered at the Radiation Center through a project tasking system. This allows for excellent quality control for the work which is performed from the point of initial contact, through the reactor application, project report generation and financial accounting. For the current fiscal year, FY95, the total cost of the reactor sharing program, including Radiation Center contributions, was $66,323.20 of which $40,000.00 was supplied by the DOE Reactor Sharing Program. The details of individual project costs is given in Table 1. The work performed for the individual projects are described in the brief work descriptions given in Table 2

  4. ARM Climate Research Facility Annual Report 2005

    Energy Technology Data Exchange (ETDEWEB)

    J. Voyles

    2005-12-31

    Through the ARM Program, the DOE funded the development of several highly instrumented ground stations for studying cloud formation processes and their influence on radiative transfer, and for measuring other parameters that determine the radiative properties of the atmosphere. This scientific infrastructure, and resultant data archive, is a valuable national and international asset for advancing scientific knowledge of Earth systems. In fiscal year (FY) 2003, the DOE designated ARM sites as a national scientific user facility: the ARM Climate Research (ACRF). The ACRF has enormous potential to contribute to a wide range interdisciplinary science in areas such as meteorology, atmospheric aerosols, hydrology, biogeochemical cycling, and satellite validation, to name only a few.

  5. General considerations for neutron capture therapy at a reactor facility

    International Nuclear Information System (INIS)

    Binney, S.E.

    2001-01-01

    In addition to neutron beam intensity and quality, there are also a number of other significant criteria related to a nuclear reactor that contribute to a successful neutron capture therapy (NCT) facility. These criteria are classified into four main categories: Nuclear design factors, facility management and operations factors, facility resources, and non-technical factors. Important factors to consider are given for each of these categories. In addition to an adequate neutron beam intensity and quality, key requirements for a successful neutron capture therapy facility include necessary finances to construct or convert a facility for NCT, a capable medical staff to perform the NCT, and the administrative support for the facility. The absence of any one of these four factors seriously jeopardizes the overall probability of success of the facility. Thus nuclear reactor facility management considering becoming involved in neutron capture therapy, should it be proven clinically successful, should take all these factors into consideration. (author)

  6. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  7. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    International Nuclear Information System (INIS)

    1995-01-01

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

  8. Annual report to the Laser Facility Committee 1986

    International Nuclear Information System (INIS)

    1986-01-01

    This paper is the annual report of the Science and Engineering Research Council, research and development work carried out at the Central Laser Facility, Rutherford Laboratory, United Kingdom, 1985/6. Part I contains the technical details of the studies of the High Power Laser scientific programme and Laser Support Facility, as well as the Laser Research and Development investigations. Part II concerns the application of UV lasers to microcircuit fabrication. (UK)

  9. Brookhaven Reactor Experiment Control Facility, a distributed function computer network

    International Nuclear Information System (INIS)

    Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

    1975-11-01

    A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

  10. 20. Annual report. OECD Halden reactor project. 1979

    International Nuclear Information System (INIS)

    1981-01-01

    This is the Twentieth Annual Report on the OECD Halden Reactor Project, describing activities during 1979, the first year of the 1979-1981 Halden Agreement. Research work at the project is focussed on three areas: 1) In-core behaviour of reactor fuel, particularly reliability and safety aspects, which is studied through irradiation of test fuel elements. 2) Prediction, surveillance and control of fuel and core performance, for which models of fuel and core behaviour are developed. 3) Applications of process computers to power plant control, for which prototype software systems and hardware arrangements are developed

  11. Progress report concerning safety research for nuclear reactor facilities

    International Nuclear Information System (INIS)

    1978-01-01

    Examination and evaluation of safety research results for nuclear reactor facilities have been performed, as more than a year has elapsed since the plan had been initiated in April, 1976, by the special sub-committee for the safety of nuclear reactor facilities. The research is carried out by being divided roughly into 7 items, and seems to be steadily proceeding, though it does not yet reach the target. The above 7 items include researches for (1) criticality accident, (2) loss of coolant accident, (3) safety for light water reactor fuel, (4) construction safety for reactor facilities, (5) reduction of release of radioactive material, (6) safety evaluation based on the probability theory for reactor facilities, and (7) aseismatic measures for reactor facilities. With discussions on the progress and the results of the research this time, research on the behaviour on fuel in abnormal transients including in-core and out-core experiments has been added to the third item, deleting the power-cooling mismatch experiment in Nuclear Safety Research Reactor of JAERI. Also it has been decided to add two research to the seventh item, namely measured data collection, classification and analysis, and probability assessment of failures due to an earthquake. For these 7 items, the report describes the concrete contents of research to be performed in fiscal years of 1977 and 1978, by discussing on most rational and suitable contents conceivable at present. (Wakatsuki, Y.)

  12. Twenty-ninth annual progress report of the Pennsylvania State University Breazeale Nuclear Reactor, July 1, 1983-June 30, 1984

    International Nuclear Information System (INIS)

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

    1984-07-01

    The twenty-ninth annual progress report of the operation of the Pennsylvania State University Breazeale Reactor is submitted in accordance with the requirements of Contract DE-AC02-76ER03409 with the United States Department of Energy. This report also provides the University administration with a summary of the operation of the facility for the past year

  13. Reactor cold neutron source facility, the first in Japan

    International Nuclear Information System (INIS)

    Utsuro, Masahiko; Maeda, Yutaka; Kawai, Takeshi; Tashiro, Tameyoshi; Sakakibara, Shoji; Katada, Minoru.

    1986-01-01

    In the Research Reactor Institute, Kyoto University, the first cold neutron source facility for the reactor in Japan was installed, and various tests are carried out outside the reactor. Nippon Sanso K.K. had manufactured it. After the prescribed tests outside the reactor, this facility will be installed soon in the reactor, and its outline is described on this occasion. Cold neutrons are those having very small energy by being cooled to about-250 deg C. Since the wavelength of the material waves of cold neutrons is long, and their energy is small, they are very advantageous as an experimental means for clarifying the structure of living body molecules and polymers, the atom configuration in alloys, and atomic and molecular movements by neutron scattering and neutron diffraction. The basic principle of the cold neutron source facility is to irradiate thermal neutrons on a cold moderator kept around 20 K, and to moderate and cool the neutrons by nuclear scattering to convert to cold neutrons. The preparatory research on cold neutrons and hydrogen liquefaction, the basic design to put the cold neutron source facility in the graphite moderator facility, the safety countermeasures, the manufacture and quality control, the operation outside the reactor and the performance are reported. The cold neutron source facility comprises a cold moderator tank and other main parts, a deuterium gas tank, a helium refrigerator and instrumentation. (Kako, I.)

  14. Fast reactor programme. Annual progress report 1982

    International Nuclear Information System (INIS)

    Plakman, J.C.

    1984-01-01

    The status of recent fast-capture cross sections for important fission-product nuclides has been reviewed; an intercomparison of evaluations for Eu-isotopes has been made and corrections have been applied to recent reported evaluations of neutron capture cross sections for Pd isotopes. An outline of the evaluation procedure for the nuclides sup(58g)Co and sup(58m)Co is given. The evaluation of the cover-gas nuclides has been completed with additional results for 36 Ar and 38 Ar. Some results of the latest fuel failure experiments under simulated reduced coolant conditions, the so-called SHOT experiments, are given. The first irradiation experiments with the prototype irradiation facility HFR-TOP 01 are described. Neutron flux calculations have been performed to determine the dimensions of a flux depression plate to achieve a symmetric flux distribution inside the fuel during pre-irradiation. The creep investigations on various heats and welded joints of DIN 1.4948 have been finished; the main findings are reported. A first project on the low-cycle fatigue behaviour of DIN 1.4948 has been completed. A three-dimensional finite element analysis has been performed on compact tension test specimen having a curved crack-front due to crack-tunneling. A code version of the VITESSE computer code has been developed to predict the thermohydraulic behaviour of distorted bundle geometries. Results from the LDA measuring programmes in the different test sections with respect to the secondary flow velocities are reported. Noise measurements in an unblocked 60 deg. reference bundle have been performed. (Auth.)

  15. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

    Brynda, W.J.; Powell, R.W.

    1978-08-01

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned critical facilities be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  16. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  17. Facility with a nuclear district heating reactor

    International Nuclear Information System (INIS)

    Straub, H.

    1988-01-01

    The district heating reactor has a pressure vessel which contains the reactor core and at least one coolant conducting primary heat carrier surrounded by a heat sink. The pressure vessel has two walls with a space between them. This space is connected with a container which contains air as heat isolating medium and water as heat conducting medium. During the normal reactor operation the space is filled by air from the container with the aid of a blower, whereas in the case of a break-down of the cooling system it is filled by water which flows out of the container by gravity after the blower has been switched off. The after-heat, generated in the reactor core during cooling break-down, is removed into the heat sink surrounding the pressure vessel in a safe and simple way. 6 figs

  18. Coolant cleaning facility for nuclear reactor

    International Nuclear Information System (INIS)

    Kuboniwa, Takao; Konno, Yasuhiro; Kumaya, Shin; Osumi, Katsumi.

    1982-01-01

    Purpose: To remove cation of radioactive cobalt 60 produced in a reactor water during the ordinary operation of the reactor and chlorine when sea water is leaked in a condenser as well as to suppress an increase in iron clad containing radioactive cobalt 60 in the reactor water when the reactor is stopped. Constitution: A large flow rate high temperature cleaning system having an electromagnetic filter capable of removing radioactive substance in a reactor water, a low temperature cleaning system having a desalting unit using ion exchanger resin, a turbidity meter for measuring the turbidity of the reactor water and a conductivity meter for measuring the conductivity are provided. Further, flow rate control means are provided in the high and low temperature cleaning systems. The flow rate control means of the high temperature cleaning system is controlled by a measured signal of the turbidity meter, and the flow rete control means of the low temperature cleaning system is controlled by the measured signal of the conductivity meter. (Aizawa, K.)

  19. Discussion of the use of the Dragon reactor as a facility for integral reactor physics experiments

    Energy Technology Data Exchange (ETDEWEB)

    Gutmann, H

    1972-06-05

    The purpose and use of the Dragon Reactor Experiment (DRE) has changed considerably during the years of its operation. The original purpose was to show that the principle of a High Temperature Reactor is sound and demonstrate its operation. After this achievement, the purpose of the Dragon reactor changed to the use as a fuel testing facility. During recent years, a new use of the DRE has been added to its use as a fuel testing facility, namely Fuel Element Design Testing. The current report covers reactor physics experiments aspects.

  20. Coupling of AST-500 heating reactors with desalination facilities

    International Nuclear Information System (INIS)

    Kourachenkov, A.V.

    1998-01-01

    The general issues regarding NHR and desalination facility joint operation for potable water production are briefly considered. AST-500 reactor plant and DOU GTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. Similarity of NHR operation for a heating grid and a desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author)

  1. Coupling of AST-500 heating reactors with desalination facilities

    International Nuclear Information System (INIS)

    Gureyeva, L.V.; Egorov, V.V.; Podberezniy, V.L.

    1997-01-01

    The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab

  2. Coupling of AST-500 heating reactors with desalination facilities

    Energy Technology Data Exchange (ETDEWEB)

    Gureyeva, L V; Egorov, V V [OKBM, Nizhny Novgorod (Russian Federation); Podberezniy, V L [Scientific Research Inst. of Machine Building, Ekaterinburg (Russian Federation)

    1997-09-01

    The general issues regarding the joint operation of a NHR and a desalination facility for potable water production are briefly considered. The AST-500 reactor plant and the DOUGTPA-type evaporating desalination facilities, both relying on proven technology and solid experience of construction and operation, are taken as a basis for the design of a large-output nuclear desalination complex. Its main design characteristics are given. The similarity of NHR operation for heating grid and desalination facility in respect of reactor plant operating conditions and power regulation principles is pointed out. The issues of nuclear desalination complexes composition are discussed briefly as well. (author). 2 refs, 1 fig., 1 tab.

  3. Design requirements for new nuclear reactor facilities in Canada

    International Nuclear Information System (INIS)

    Shim, S.; Ohn, M.; Harwood, C.

    2012-01-01

    The Canadian Nuclear Safety Commission (CNSC) has been establishing the regulatory framework for the efficient and effective licensing of new nuclear reactor facilities. This regulatory framework includes the documentation of the requirements for the design and safety analysis of new nuclear reactor facilities, regardless of size. For this purpose, the CNSC has published the design and safety analysis requirements in the following two sets of regulatory documents: 1. RD-337, Design of New Nuclear Power Plants and RD-310, Safety Analysis for Nuclear Power Plants; and 2. RD-367, Design of Small Reactor Facilities and RD-308, Deterministic Safety Analysis for Small Reactor Facilities. These regulatory documents have been modernized to document past practices and experience and to be consistent with national and international standards. These regulatory documents provide the requirements for the design and safety analysis at a high level presented in a hierarchical structure. These documents were developed in a technology neutral approach so that they can be applicable for a wide variety of water cooled reactor facilities. This paper highlights two particular aspects of these regulatory documents: The use of a graded approach to make the documents applicable for a wide variety of nuclear reactor facilities including nuclear power plants (NPPs) and small reactor facilities; and, Design requirements that are new and different from past Canadian practices. Finally, this paper presents some of the proposed changes in RD-337 to implement specific details of the recommendations of the CNSC Fukushima Task Force Report. Major changes were not needed as the 2008 version of RD-337 already contained requirements to address most of the lessons learned from the Fukushima event of March 2011. (author)

  4. Intense neutron irradiation facility for fusion reactor materials

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio; Kato, Yoshio; Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Technical R and D of d-Li stripping type neutron irradiation facilities for development of fusion reactor materials was carried out in Fusion Materials Irradiation Test Facility (FMIT) project and Energy Selective Neutron Irradiation Test Facility (ESNIT) program. Conceptual design activity (CDA) of International Fusion Materials Irradiation Facility (IFMIF), of which concept is an advanced version of FMIT and ESNIT concepts, are being performed. Progress of users` requirements and characteristics of irradiation fields in such neutron irradiation facilities, and outline of baseline conceptual design of IFMIF were described. (author)

  5. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  6. 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

  7. Hanford Facility Annual Dangerous Waste Report Calendar Year 2002

    International Nuclear Information System (INIS)

    FR-EEMAN, D.A.

    2003-01-01

    Hanford CY 2002 dangerous waste generation and management forms. The Hanford Facility Annual Dangerous Waste Report (ADWR) is prepared to meet the requirements of Washington Administrative Code Sections 173-303-220, Generator Reporting, and 173-303-390, Facility Reporting. In addition, the ADWR is required to meet Hanford Facility RCRA Permit Condition I.E.22, Annual Reporting. The ADWR provides summary information on dangerous waste generation and management activities for the Calendar Year for the Hanford Facility EPA ID number assigned to the Department of Energy for RCRA regulated waste, as well as Washington State only designated waste and radioactive mixed waste. The Solid Waste Information and Tracking System (SWITS) database is utilized to collect and compile the large array of data needed for preparation of this report. Information includes details of waste generated on the Hanford Facility, waste generated offsite and sent to Hanford for management, and other waste management activities conducted at Hanford, including treatment, storage, and disposal. Report details consist of waste descriptions and weights, waste codes and designations, and waste handling codes. In addition, for waste shipped to Hanford for treatment and/or disposal, information on manifest numbers, the waste transporter, the waste receiving facility, and the original waste generators are included. In addition to paper copies, electronic copies of the report are also transmitted to the regulatory agency

  8. Emergency water supply facility for nuclear reactor

    International Nuclear Information System (INIS)

    Karasawa, Toru

    1998-01-01

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

  9. 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

  10. Annual report of department of research reactor, 1992

    International Nuclear Information System (INIS)

    1993-12-01

    The department of research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1992 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as well as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  11. Annual Summary of the Integrated Disposal Facility Performance Assessment 2012

    Energy Technology Data Exchange (ETDEWEB)

    Khaleel, R. [INTERA, Austin, TX (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2012-12-27

    An annual summary of the adequacy of the Hanford Immobilized Low-Activity Waste (ILAW) Performance Assessment (PA) is required each year (DOE O 435.1 Chg 1,1 DOE M 435.1-1 Chg 1;2 and DOE/ORP-2000-013). The most recently approved PA is DOE/ORP-2000-24.4 The ILAW PA evaluated the adequacy of the ILAW disposal facility, now referred to as the Integrated Disposal Facility (IDF), for the safe disposal of vitrified Hanford Site tank waste.

  12. Standard irradiation facilities for use in TRIGA reactors

    International Nuclear Information System (INIS)

    Kolbasov, B.N.; Luse, R.A.

    1972-01-01

    The standard neutron irradiation facility (SNIP) was developed under IAEA and FAO co-ordinated research program for the standardization of neutron irradiation facilities for radiobiological research, resulting in the possibility to use fast neutrons from pool-type reactors for radiobiological studies. The studies include irradiation of seeds for crop improvement, of Drosophila for genetic studies, and of microorganisms for developing industrially useful mutants, as well as fundamental studies in radiation biology. The facilities, located in the six pool-type reactors (in Austria, Bulgaria, India, Philippines, Thailand and Taiwan), have been calibrated and utilized to compare the response to fast neutrons of barley seeds (variety Himalaya CI 000620) which were selected as a standard biological monitor by which to estimate neutron fluxes in different reactors. These comparative irradiation studies showed excellent agreement and reproducibility

  13. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  14. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  15. Neutron beam facilities at the replacement research reactor

    International Nuclear Information System (INIS)

    Kennedy, S.

    1999-01-01

    Full text: On September 3rd 1997 the Australian Federal Government announced their decision to replace the HIFAR research reactor by 2005. The proposed reactor will be a multipurpose reactor with improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The neutron beam facilities are intended to cater for Australian scientific needs well into the 21st century. In the first stage of planning the neutron Beam Facilities at the replacement reactor, a Consultative Group was formed (BFCG) to determine the scientific capabilities of the new facility. Members of the group were drawn from academia, industry and government research laboratories. The BFCG submitted their report in April 1998, outlining the scientific priorities to be addressed. Cold and hot neutron sources are to be included, and cold and thermal neutron guides will be used to position most of the instruments in a neutron guide hall outside the reactor confinement building. In 2005 it is planned to have eight instruments installed with a further three to be developed by 2010, and seven spare instrument positions for development of new instruments over the life of the reactor. A beam facilities technical group (BFTG) was then formed to prepare the engineering specifications for the tendering process. The group consisted of some members of the BFCG, several scientists and engineers from ANSTO, and scientists from leading neutron scattering centres in Europe, USA and Japan. The BFTG looked in detail at the key components of the facility such as the thermal, cold and hot neutron sources, neutron collimators, neutron beam guides and overall requirements for the neutron guide hall. The report of the BFTG, completed in August 1998, was incorporated into the draft specifications for the reactor project, which were distributed to potential reactor vendors. An assessment of the first stage of reactor vendor submissions was completed in

  16. Proposal for a seismic facility for reactor safety research

    International Nuclear Information System (INIS)

    Anderson, C.A.; Dove, R.C.; Rhorer, R.L.

    1976-07-01

    Certain problem areas in the seismic analysis and design of nuclear reactors are enumerated and the way in which an experimental program might contribute to each area is examined. The use of seismic simulation testing receives particular attention, especially with regard to the verification of structural response analysis. The importance of scale modeling used in conjunction with seismic simulation is also stressed. The capabilities of existing seismic simulators are summarized, and a proposed facility is described which would considerably extend the ability to conduct, with confidence, confirmatory experiments on the behavior of reactor components when subjected to seismic excitation. Particular applications to gas-cooled and other reactor types are described

  17. Saint-Laurent-des-Eaux nuclear facilities. 2009 annual report

    International Nuclear Information System (INIS)

    2010-01-01

    This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of Saint-Laurent-des-Eaux, and then the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities, administrative procedures carried out in 2009), incidents and accidents registered in 2009, radioactive and chemical effluents released by the facilities in the environment, other pollutions, management of radioactive wastes, and, finally, the actions carried out in the domain of transparency and public information. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

  18. Gamma ray thermometrical facility for nuclear reactor

    International Nuclear Information System (INIS)

    Smith, R.D.; Regazzoni, Pierre.

    1981-01-01

    This invention concerns a gamma ray thermometer for nuclear reactors, fitted with a thermal bridge for use as a centring device. In accordance with the invention, an elastic device fills all the annular space between the gamma thermometer and the orifice through which the thermometer is introduced. This elastic device has the two-fold role of providing a thermal bridge at the gamma thermometer location suitable as a heat well, and of acting as a device for centring the thermometer in the orifice into which it has been introduced [fr

  19. Quasielastic neutron scattering facility at Dhruva reactor

    International Nuclear Information System (INIS)

    Mukhopadhyay, R.; Mitra, S.; Paranjpe, S.K.; Dasannacharya, B.A.

    2001-01-01

    Quasi-elastic neutron scattering is a powerful experimental tool for studying the various dynamical motions in solids and liquids. In this paper, we have described the salient features of the quasi-elastic neutron spectrometer in operation at Dhruva reactor at Trombay, India. The design criteria have been such as to maximise the throughput by various means like closer approach to the source, focusing a larger beam on to a sample, and Multi-Angle Reflecting X-tal mode of energy analysis. Some results of molecular motions from recently studied systems using this spectrometer are also reported

  20. Pressure suppression facility for reactor container

    International Nuclear Information System (INIS)

    Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

    1993-01-01

    In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

  1. Jordan Research and Training Reactor (JRTR) Utilization Facilities

    International Nuclear Information System (INIS)

    Xoubi, N.

    2013-01-01

    Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

  2. 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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  4. Abnormal Events for Reactor System and Facilities in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ho Young; Lee, B. H.; Lee, M.; Kang, I. H.; Lee, U. G.; Sin, H. C.; Park, C. Y.; Song, B. S.; Lee, S. H.; Han, J. S

    2006-12-15

    This report gathers abnormal events related to reactor system and facilities of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide.

  5. The neutron beam facility at the Australian replacement research reactor

    International Nuclear Information System (INIS)

    Hunter, B.; Kennedy, S.

    1999-01-01

    Full text: The Australian federal government gave ANSTO final approval to build a research reactor to replace HIFAR on August 25th 1999. The replacement reactor is to be a multipurpose reactor with a thermal neutron flux of 3 x 10 14 n.cm -2 .s -1 and having improved capabilities for neutron beam research and for the production of radioisotopes for pharmaceutical, scientific and industrial use. The replacement reactor will commence operation in 2005 and will cater for Australian scientific, industrial and medical needs well into the 21st century. The scientific capabilities of the neutron beams at the replacement reactor are being developed in consultation with representatives from academia, industry and government research laboratories to provide a facility for condensed matter research in physics, chemistry, materials science, life sciences, engineering and earth sciences. Cold, thermal and hot neutron sources are to be installed, and neutron guides will be used to position most of the neutron beam instruments in a neutron guide hall outside the reactor confinement building. Eight instruments are planned for 2005, with a further three to be developed by 2010. A conceptual layout for the neutron beam facility is presented including the location of the planned suite of neutron beam instruments. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by an accredited reactor builder in a turnkey contract. Tenders have been called for December 1999, with selection of contractor planned by June 2000. The neutron beam instruments will be developed by ANSTO and other contracted organisations in consultation with the user community and interested overseas scientists. The facility will be based, as far as possible, around a neutron guide hall that is be served by three thermal and three cold neutron guides. Efficient transportation of thermal and cold neutrons to the guide hall requires the use of modern super

  6. Advanced Test Reactor National Scientific User Facility Partnerships

    International Nuclear Information System (INIS)

    Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

    2012-01-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  7. Environmental assessment for the deactivation of the N Reactor facilities

    International Nuclear Information System (INIS)

    1995-05-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. As required by the National Environmental Policy Act of 1969 (NEPA), this EA complies with Title 40, Code of Federal Regulations (CFR), parts 1500--1508, ''Regulations for Implementing the Procedural Provisions of NEPA. '' It also implements the ''National Environmental Policy Act; Implementing Procedures and Guidelines'' (10 CFR 1021)

  8. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    International Nuclear Information System (INIS)

    Stephens, S.V.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended

  9. Ventilation safety of facilities comprising nuclear reactors

    International Nuclear Information System (INIS)

    Guirlet, J.

    1982-01-01

    The reliability of the ventilation is one of the most important aspects in the prevention of the nuisances that a nuclear installation can provide, since the ventilation is located at the last barrier. A certain number of essential points have been recalled here. But it is necessary to bear in mind other requirements such as the limitation in the number of crossovers, the answers to be found should the system fail, the need to show that ventilation systems do not in themselves bring other nuisances such as noise, irradiation or contamination hazards, likelyhood of recycling the contamination, vibrations, fire. Finally, it is absolutely essential, right from the project stage, that the design ensures that very good accessibility, very easy dismantling and handling, as well as all the facilities needed to make sure of the initial and periodic tests, are guaranteed [fr

  10. Surveys of research projects concerning nuclear facility safety, financed by the Federal Ministry for the Environment, Nature Protection and Reactor Safety, 1989. (14. annual report on SR-projects)

    International Nuclear Information System (INIS)

    1990-11-01

    Each progress report is a collection of individual reports, categorized by subject matter. They are a documentation of the contractor's progress, rendered by themselves on standardized forms, published, for the sake of general information on progress made in investigations concerning reactor safety, by the project attendance department of the GRS. The individual reports have serial numbers. Each report includes particulars of the objective, work carried out, results obtained and plans for project continuation. (orig.) [de

  11. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  12. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  13. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    Kim, S.

    2003-01-01

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  14. Manufacture and installation of reactor auxiliary facilities for advanced thermal prototype reactor 'Fugen'

    International Nuclear Information System (INIS)

    Kawahara, Toshio; Matsushita, Tadashi

    1977-01-01

    The facilities of reactor auxiliary systems for the advanced thermal prtotype reactor ''Fugen'' were manufactured in factories since 1972, and the installation at the site began in November, 1974. It was almost completed in March, 1977, except a part of the tests and inspections, therefore the outline of the works is reported. The ATR ''Fugen'' is a heavy water-moderated, boiling light water reactor, and its reactor auxiliary systems comprise mainly the facilities for handling heavy water, such as heavy water cooling system, heavy water cleaning system, poison supplying system, helium circulating system, helium cleaning system, and carbon dioxide system. The poison supplying system supplies liquid poison to the heavy water cooling system to absorb excess reactivity in the initial reactor core. The helium circulating system covers heavy water surface with helium to prevent the deterioration of heavy water and maintains heavy water level by pressure difference. The carbon dioxide system flows highly pure CO 2 gas in the space of pressure tubes and carandria tubes, and provides thermal shielding. The design, manufacture and installation of the facilities of reactor auxiliary systems, and the helium leak test, synthetic pressure test and total cleaning are explained. (Kako, I.)

  15. A neutron radiography facility on the IRT-2000 reactor

    International Nuclear Information System (INIS)

    Khadduri, I.Y.

    1976-01-01

    A neutron radiography facility has been constructed on the thermal neutron channel of the IRT-2000 reactor. A collimated thermal neutron beam exposure area of 10 cm diameter is obtained with an L/D ratio of 48.8. The film used is cellulose nitrate coated with lithium tetraborate which is insensitive to gamma and beta radiation. Some pictures with good contrast and resolution have been obtained. Pictures of parts of an IRT-2000 reactor fuel pin have also been recorded. (orig) [de

  16. Research reactor facilities, recent developments at Imperial College, London

    International Nuclear Information System (INIS)

    Franklin, S.J.; Goddard, A.J.H.; O Connell, J.

    1998-01-01

    The 100 kW CONSORT pool-type reactor is now the only Research Reactor in the UK. Because of its strategic importance, Imperial College is continuing and accelerating a programme of refurbishment of the control system, and planning for a further fuel charge. These plans are described and the progress to date discussed. To this end, a description of the enhanced Safety Case being written is provided here and refueling plans discussed. The current range of facilities available is described, and future plans highlighted. (author)

  17. Low temperature irradiation facility at Kyoto University Reactor (KUR)

    International Nuclear Information System (INIS)

    Atobe, Kozo; Okada, Moritami; Yoshida, Hiroyuki; Kodaka, Hisao; Miyata, Kiyomi.

    1977-01-01

    A new refrigeration system has been substituted to the low temperature irradiation facility at KUR instead of the previous one, since April in 1975. The model 1204 CTi He liquifier was designed to be modified for the refrigerator with the capacity of 30 watts at 10 K. The refrigeration capacity of 38 watts at 10 K was defined using a special cryostat and transfer-tubes, and the lowest temperature of about 18 K was measured using the irradiation loop without reactor operation. The reconstructed facility enables us to hold the many specimens simultaneously in the sample chamber of the irradiation loop at about 25 K during reactor operation of 5 MW. The irradiation dose has been reached about 6.6 x 10 16 n sub(f)/cm 2 and 6.1 x 10 17 n sub(th)/cm 2 with the normal reactor operation cycle of up to 77 hours. The stable operation condition of the machine and the special safety system for the refrigeration system enable us to maintain easily the facility with a constant operation condition for such a long time irradiation. Many kinds of low temperature neutron irradiation experiments are carried out using the facility, which techniques are partially reported. (auth.)

  18. Neutron beam facilities at the Australian Replacement Research Reactor

    International Nuclear Information System (INIS)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett

    2001-01-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10 14 n/cm 2 /sec and a liquid D 2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  19. Reactor engineering department annual report. April 1, 1993-March 31, 1994

    International Nuclear Information System (INIS)

    1994-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1993 (April 1, 1993-March 31, 1994). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees organized by the Department are also summarized in this report. (author)

  20. RB Research nuclear reactor, Annual report for 2005

    International Nuclear Information System (INIS)

    Milosevic, M.; Dasic, N.; Ljubenov, V.; Pesic, M.; Nikolic, D; Jevremovic, M.; Minic, D.

    2006-01-01

    Report on RB reactor operation during 2005 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2005

  1. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  2. UCN-VCN facility and experiments in Kyoto University Reactor

    International Nuclear Information System (INIS)

    Kawabata, Yuji; Okumura, Kiyoshi; Utsuro, Masahiko

    1993-01-01

    An ultracold and very cold neutron facility was installed in Kyoto University Reactor (KUR). The facility consists of a very cold neutron (VCN) guide tube, a VCN bender, a supermirror neutron turbine and experimental equipments with ultracold neutrons (UCN). The properties of each equipments are presented. UCN is generated by a supermirror neutron turbine combined with the cold neutron source operated with liquid deuterium, and the UCN output spectrum was measured by the time-of-flight method. A gravity analyzer for high resolution spectroscopy and a neutron bottle for decay experiments are now developing as the UCN research in KUR. (author)

  3. The reactor and cold neutron research facility at NIST

    Energy Technology Data Exchange (ETDEWEB)

    Prask, H J; Rowe, J M [Reactor Radiation Division, National Institute of Standards and Technology, Gaithersburg, MD (United States)

    1992-07-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D{sub 2}O/H{sub 2}O ice cold source. (author)

  4. THE WHITE SANDS MISSILE RANGE PULSED REACTOR FACILITY, MAY 1963

    Energy Technology Data Exchange (ETDEWEB)

    Long, Robert L.; Boor, R. A.; Cole, W. M.; Elder, G. E.

    1963-05-15

    A brief statement of the mission of the White Sands Missile Range Nuclear Effects Laboratory is given. The new Nuclear Effects Laboratory Facility is described. This facility consists of two buildings-a laboratory and a reactor building. The White Sands Missile Range bare critical assembly, designated as the MoLLY-G, is described. The MoLLY-G, an unreflected, unmoderated right circular cylinder of uranium-molybdenum alloy designed for pulsed operation, will have a maximum burst capability of approximately 2 x 10/sup 17/ fissions with a burst width of 50 microseconds. The reactor construction and operating procedures are described. As designed, the MoLLY-G will provide an intense source of pulsed neutron and gamma radiation for a great variety of experimental and test arrangements. (auth)

  5. The reactor and cold neutron research facility at NIST

    International Nuclear Information System (INIS)

    Prask, H.J.; Rowe, J.M.

    1992-01-01

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D 2 O/H 2 O ice cold source. (author)

  6. 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

  7. 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

  8. Atmospheric Radiation Measurement Climate Research Facility (ACRF) Annual Report 2007

    Energy Technology Data Exchange (ETDEWEB)

    LR Roeder

    2007-12-01

    This annual report describes the purpose and structure of the program, and presents key accomplishments in 2007. Notable achievements include: • Successful review of the ACRF as a user facility by the DOE Biological and Environmental Research Advisory Committee. The subcommittee reinforced the importance of the scientific impacts of this facility, and its value for the international research community. • Leadership of the Cloud Land Surface Interaction Campaign. This multi-agency, interdisciplinary field campaign involved enhanced surface instrumentation at the ACRF Southern Great Plains site and, in concert with the Cumulus Humilis Aerosol Processing Study sponsored by the DOE Atmospheric Science Program, coordination of nine aircraft through the ARM Aerial Vehicles Program. • Successful deployment of the ARM Mobile Facility in Germany, including hosting nearly a dozen guest instruments and drawing almost 5000 visitors to the site. • Key advancements in the representation of radiative transfer in weather forecast models from the European Centre for Medium-Range Weather Forecasts. • Development of several new enhanced data sets, ranging from best estimate surface radiation measurements from multiple sensors at all ACRF sites to the extension of time-height cloud occurrence profiles to Niamey, Niger, Africa. • Publication of three research papers in a single issue (February 2007) of the Bulletin of the American Meteorological Society.

  9. RB Research nuclear reactor, Annual report for 2004

    International Nuclear Information System (INIS)

    Dasic, N.; Pesic, M.; Nikolic, D; Jevremovic, M.; Eskirovic, B.

    2005-02-01

    Report on RB reactor operation during 2004 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. It contains data about reactor operation during previous 8 years. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation, Annex 1. contains data about heavy water degradation, and Annex 2 is the certificate about the crane bridge in the reactor hall

  10. Japan Atomic Energy Research Institute, Reactor Engineering Division annual report

    International Nuclear Information System (INIS)

    1981-09-01

    Research activities in the Division of Reactor Engineering in fiscal 1980 are described. The work of the Division is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor and fusion reactor, and development of Liquid Metal Fast Breeder Reactor carried out by Power Reactor and Nuclear Fuel Development Corporation. Contents of the report are achievements in fields such as nuclear data and group constants, theoretical method and code development, integral experiment and analysis, shielding, reactor and nuclear instrumentation, reactor control and diagnosis, and fusion reactor technology, and activities of the Committee on Reactor Physics. (author)

  11. The proposed cold neutron irradiation facility at the Breazeale reactor

    International Nuclear Information System (INIS)

    Dimeo, R. M.; Sokol, P. E.; Carpenter, J. M.

    1997-01-01

    We discuss the design considerations of a Cold Neutron Irradiation Facility (CNIF) originally to have been installed at the Penn State Breazeale Reactor (PSBR). The goal of this project was to study the effects of radiation-induced damage to cryogenic moderators and, in particular, solid methane. This work evolved through the design stage undergoing a full safety analysis and received tentative approval from the PSBR Safeguards Committee but was discontinued due to budgetary constraints. (auth)

  12. Facility for in-reactor creep testing of fuel cladding

    International Nuclear Information System (INIS)

    Kohn, E.; Wright, M.G.

    1976-11-01

    A biaxial stress creep test facility has been designed and developed for operation in the WR-1 reactor. This report outlines the rationale for its design and describes its construction and the operating experience with it. The equipment is optimized for the determination of creep data on CANDU fuel cladding. Typical results from Zr-2.5 wt% Nb fuel cladding are used to illustrate the accuracy and reliability obtained. (author)

  13. Training and research reactor facility longevity extension program

    International Nuclear Information System (INIS)

    Carriveau, G.W.

    1991-01-01

    Since 1943, over 550 training and research reactors have been in operation. According to statistics from the International Atomic Energy Agency, ∼325 training and research reactors are currently in service. This total includes a wide variety of designs covering a range of power and research capabilities located virtually around the world. A program has been established at General Atomics (GA) that is dedicated to the support of extended longevity of training and research reactor facilities. Aspects of this program include the following: (1) new instrumentation and control systems; (2) improved and upgraded nuclear monitoring and control channels; (3) facility testing, repair and upgrade services that include (a) pool or tank integrity, (b) cooling system, and (c) water purification system; (4) fuel element testing procedures and replacement; (5) control rod drive rebuilding and upgrades; (6) control and monitoring system calibration and repair service; (7) training services, including reactor operations, maintenance, instrumentation calibration, and repair; and (8) expanded or new uses such as neutron radiography and autoradiography, isotope production, nuclear medicine, activation analysis, and material properties modification

  14. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    participating in research related to the development of the GFR. These are Euratom (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, Euratom (including the United Kingdom and Switzerland), France, and Japan have active research activities with respect to the GFR. The research includes GFR design and safety, and fuels/in-core materials/fuel cycle projects. This report outlines the current design status of the GFR, and includes work done in the areas mentioned above for this fiscal year. In addition, this report fulfills the Level 2 milestones, ''Complete annual status report on GFR reactor design'', and ''Complete annual status report on pre-conceptual GFR reactor designs'' in work package GI0401K01. GFR funding for FY05 included FY04 carryover funds, and was comprised of multiple tasks. These tasks involved a consortium of national laboratories and universities, including the Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Oak Ridge National Laboratory (ORNL), Auburn University (AU), Idaho State University (ISU), and the University of Wisconsin-Madison (UW-M). The total funding for FY05 was $1000K, with FY04 carryover of $174K. The cost breakdown can be seen in Table 1.

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  16. The application of SQUG to non-reactor facilities

    International Nuclear Information System (INIS)

    Hoskins, R.S.

    1993-01-01

    DOE Order 6430.1A mandates that facilities be designed and constructed to withstand Natural Phenomena Hazards in accordance with their hazard level. DOE has a program in progress to evaluate and then upgrade many of their existing medium and high hazard facilities where release of hazardous materials to the environment is a concern. This paper addresses a useful methodology which has been applied by SRS to evaluate and qualify equipment to withstand the ravenousness of earthquakes. The Seismic Qualification Utility Group was formed by a group of Electric Power Utilities whose Nuclear Power Plants predated the 10CFR50 Environmental design requirements to develop a methodology to evaluate and upgrade their operating plants against seismic events in answer to NRC generic letter USI-A46. SRS participated in this organization, since it operated reactors designed and constructed in the 1950's, and the application of the SQUG methodology was obvious. Nuclear Material Processing and Handling (NMPH) facilities utilize equipment similar to the nuclear industry, in fact, to industry in general. Consequently, it made sense to apply SQUG methodology to evaluate and qualify equipment in NMPH facilities against earthquakes. In order to utilize SQUG methodology, some changes are required since the goal of safe shutdown of a NMPH facility differs from a nuclear reactor, consequently, the generation of a Safe Shutdown Equipment List is modified by the requirements of each particular facility. Once the Safety Class Equipment List is developed denoting the qualification requirement for each piece of equipment, the SQUG walk downs can be performed as they are in commercial nuclear plants. SQUG methodology offers a cost effective approach for the seismic qualification of equipment at existing DOE NMPH facilities. SQUG cannot be applied to new or unique equipment since the experience database doesn't contain the needed information

  17. Radiation protection planning for decommissioning of research reactor facilities

    International Nuclear Information System (INIS)

    Jackson, Roger; Harman, Neil; Craig, David; Fecitt, Lorna; Lobach, Yuri; Gorlinskij, Juri; Kolyadin, Vyacheslav; Pavlenko, Vytali

    2008-01-01

    The MR reactor at the Russian Research Centre Kurchatov Institute (RRCKI), Moscow was a 50 MW multipurpose material testing and research reactor equipped with nine experimental loop facilities to test prototype fuel for various nuclear power reactors being developed. The reactor was shut down in 1993 and de-fuelled. The experimental loops are located in basement rooms around the reactor. The nature of the research into the characteristics of fuel design and coolant chemistry resulted in fission products and activation products in the test loop equipment. Decommissioning of the loops therefore presents a number of challenges. In addition the city of Moscow has expanded such that the RRC KI is now surrounded by housing which had to be taken into account in the radiological protection planning. This paper describes the techniques proposed to undertake the dismantling operations in order to minimise the radiation exposure to workers and members of the public. Estimates have been made of the worker doses which could be incurred during the dismantling process and the environmental impacts which could occur. These are demonstrated to be as low as reasonably achievable. The work was funded by the UK Department of Business Enterprise and Regulatory Reform (DBERR) (formerly the Department of Trade and Industry) under the Nuclear Safety Programme (NSP) set up to address nuclear safety issues in the Former Soviet Union. (author)

  18. RB research nuclear reactor, Annual report for 1989, I - III

    International Nuclear Information System (INIS)

    Stefanovic, D.; Pesic, M.; Hadimahmutovic, N.; Vranic, S.; Petronijevic, M.; Jevremovic, M.; Ilic, I.

    1989-12-01

    This report is made of three parts. Part one contains a short description of the reactor, reactor operation, incidents, status of reactor equipment and components (nuclear fuel, heavy water, reactor vessel, heavy water circulation system, electronic, electric and mechanical equipment, auxiliary systems and Vax-8250 computer). It includes dosimetry and radiation protection data, personnel and financial data. Second part of this report in concerned with maintenance of reactor components and instrumentation. Part three includes data about reactor utilization during 1989

  19. RB Research nuclear reactor, Annual report for 2007

    International Nuclear Information System (INIS)

    Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.; Sipka, Dj.

    2008-01-01

    Report on RB reactor operation during 2007 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation during 2007. Majority of measurement were related to spent fuel from the RA reactor, safety of transportation containers and verification of relevant computer codes

  20. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    International Nuclear Information System (INIS)

    Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

  1. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  2. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    International Nuclear Information System (INIS)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  3. Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

    International Nuclear Information System (INIS)

    Chatt, A.; Holzbecher, J.

    1990-01-01

    SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

  4. Operation and maintenance of the RB reactor, Annual report for 1979

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.; Petronijevic, M.

    1979-01-01

    The annual report for 1979 includes the following: utilization of the RB reactor; accident analysis; dosimetry and radiation protection; description of the reactor equipment status; reactor staff; financial data. Six Annexes to this report are concerned with: scientific program orientation; producing the new safety report; maintenance of the reactor components and equipment; purchase of new equipment; financial report; and program for training the staff of the Krsko NPP

  5. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1991

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1993-07-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1991. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1991 annual reports submitted by about 436 licensees indicated that approximately 206,732 individuals were monitored, 182,334 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.15 rem (cSv) and an average measurable dose of about 0.31 (cSv). Termination radiation exposure reports were analyzed to reveal that about 96,231 individuals completed their employment with one or more of the 436 covered licensees during 1991. Some 68,115 of these individuals terminated from power reactor facilities, and about 7,763 of them were considered to be transient workers who received an average dose of 0.52 rem (cSv)

  6. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1989

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1992-04-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC 1 licensees during the years 1969 through 1989. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC 1 licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1989 annual reports submitted by about 448 licensees indicated that approximately 216,294 individuals were monitored 111,000 of whom were monitored by nuclear power facilities. They incurred an average individual does of 0.18 rem (cSv) and an average measurable dose of 0.36 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,535 individuals completed their employment with one or more of the 448 covered licensees during 1989. Some 76,561 of these individuals terminated from power reactor facilities, and about 10, 344 of them were considered to be transient workers who received an average dose of 0.64 rem (cSv)

  7. Occupational radiation exposure at commercial nuclear power reactors and other facilities, 1988

    International Nuclear Information System (INIS)

    Raddatz, C.T.

    1991-07-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC's Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1988. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10 CFR 20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10 CFR 20.408. The 1988 annual reports submitted by about 429 licensees indicated that approximately 220,048 individuals were monitored, 113,00 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.20 rem (cSv) and an average measurable dose of 0.41 (cSv). Termination radiation exposure reports were analyzed to reveal that about 113,072 individuals completed their employment with one or more of the 429 covered licensees during 1988. Some 80,211 of these individuals terminated from power reactor facilities, and about 8,760 of them were considered to be transient workers who received an average dose of 0.27 rem (cSv). 17 refs., 11 figs., 29 tabs

  8. Reactor Engineering Department annual report (April 1, 1990 - March 31, 1991)

    International Nuclear Information System (INIS)

    1991-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1990 (April 1, 1990 - March 31, 1991). The major Department's programs promoted in the year are the assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  9. Reactor Engineering Department annual report (April 1, 1991-March 31, 1992)

    International Nuclear Information System (INIS)

    1992-08-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1991 (April 1, 1991-March 31, 1992). The major Department's programs promoted in the year are assessment of the high conversion light water reactor, the design activities of advanced reactor system and development of a high energy proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researchers on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics, technology assessment of nuclear energy and technology developments related to the reactor physics facilities. The cooperative work to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project also progressed. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  10. Reactor Engineering Department annual report (April 1, 1987 - March 31, 1988)

    International Nuclear Information System (INIS)

    1988-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988). The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC. The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics. (author)

  11. Reactor Engineering Department annual report (April 1, 1988 - March 31, 1989)

    International Nuclear Information System (INIS)

    1989-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1988 (April 1, 1988 - March 31, 1989). The Department has promoted cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and also to PNC's fast reactor project. Other major Department's programs are the assessment of the high conversion light water reactor and the design activities of advanced reactor system. Application of a high energy accelerator to the nuclear engineering is also preliminarily assessed. The report also contains the latest progress in various basic researches as nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/ diagnosis and technical developments related to the reactor physics facilities. The activities of the Research Committee on Reactor Physics are also summarized. (author)

  12. 2016 Annual Inspection and Radiological Survey Results for the Piqua, Ohio, Decommissioned Reactor Site, July 2016

    Energy Technology Data Exchange (ETDEWEB)

    Zimmerman, Brian [USDOE Office of Legacy Management, Washington, DC (United States); Miller, Michele [Navarro Research and Engineering, Oak Ridge, TN (United States)

    2016-07-01

    This report presents the findings of the annual inspection and radiological survey of the Piqua, Ohio, Decommissioned Reactor Site (site). The decommissioned nuclear power demonstration facility was inspected and surveyed on April 15, 2016. The site, located on the east bank of the Great Miami River in Piqua, Ohio, was in fair physical condition. There is no requirement for a follow-up inspection, partly because City of Piqua (City) personnel participated in a March 2016 meeting to address reoccurring safety concerns. Radiological survey results from 104 locations revealed no removable contamination. One direct beta activity reading in a floor drain on the 56-foot level (1674 disintegrations per minute [dpm]/100 square centimeters [cm2]) exceeded the minimum detectable activity (MDA). Beta activity has been detected in the past at this floor drain. The reading was well below the action level of 5000 dpm/100 cm2.

  13. RB Research nuclear reactor, Annual report for 2006

    International Nuclear Information System (INIS)

    Milosevic, M.; Ljubenov, V.; Pesic, M.; Jevremovic, M.; Minic, D.

    2007-01-01

    Report on RB reactor operation during 2006 contains 3 parts. Part one contains a brief description of the reactor, reactor operation and operational capabilities, reactor components, relevant dosimetry and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains detailed data concerned with measurements performed at the RB reactor and a number of significant results obtained

  14. 1996 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    1996-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  15. The 'MELUSINE' reactor at Grenoble, France, and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the MELUSINE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities and specialized irradiation devices (loops and capsules). The information is presented in the form of six information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities

  16. Personal neutron dosimetry at a research reactor facility

    International Nuclear Information System (INIS)

    Kamenopoulou, V.; Carinou, E.; Stamatelatos, I.E.

    2001-01-01

    Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. (author)

  17. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1980-05-01

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.) [de

  18. Applied research and service activities at the University of Missouri Research Reactor Facility (MURR)

    International Nuclear Information System (INIS)

    Alger, D.M.

    1987-01-01

    The University Of Missouri operates MURR to provide an intense source of neutron and gamma radiation for research and applications by experimenters from its four campuses and by experimenters from other universities, government and industry. The 10 MW reactor, which has been operating an average of 155 hours per week for the past eight years, produces thermal neutron fluxes up to 6-7x10 14 n/cm 2 -s in the central flux trap and beamport source fluxes of up to 1.2x10 14 n/cm 2 -s. The mission of the reactor facility, to promote research, education and service, is the same as the overall mission of the university and therefore, applied research and service supported by industrial firms have been welcomed. The university recognized after a few years of reactor operation that in order to build utilization, it would be necessary to develop in-house research programs including people, equipment and activity so that potential users could more easily and quickly obtain the results needed. Nine research areas have been developed to create a broadly based program to support the level of activity needed to justify the cost of operating the facility. Applied research and service generate financial support for about one-half of the annual budget. The applied and service programs provide strong motivation for university/industry association in addition to the income generated. (author)

  19. The Advanced Test Reactor Irradiation Facilities and Capabilities

    International Nuclear Information System (INIS)

    S. Blaine Grover; Raymond V. Furstenau

    2007-01-01

    The Advanced Test Reactor (ATR) is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. The ATR is a very versatile facility with a wide variety of experimental test capabilities for providing the environment needed in an irradiation experiment. These different capabilities include passive sealed capsule experiments, instrumented and/or temperature-controlled experiments, and pressurized water loop experiment facilities. The ATR has enhanced capabilities in experiment monitoring and control systems for instrumented and/or temperature controlled experiments. The control systems utilize feedback from thermocouples in the experiment to provide a custom blended flowing inert gas mixture to control the temperature in the experiments. Monitoring systems have also been utilized on the exhaust gas lines from the experiment to monitor different parameters, such as fission gases for fuel experiments, during irradiation. ATR's unique control system provides axial flux profiles in the experiments, unperturbed by axially positioned control components, throughout each reactor operating cycle and over the duration of test programs requiring many years of irradiation. The ATR irradiation positions vary in diameter from 1.6 cm (0.625 inches) to 12.7 cm (5.0 inches) over an active core length of 122 cm (48.0 inches). Thermal and fast neutron fluxes can be adjusted radially across the core depending on the needs of individual test programs. This paper will discuss the different irradiation capabilities available and the cost/benefit issues related to each capability. Examples of different experiments will also be discussed to demonstrate the use of the capabilities and facilities at ATR for performing irradiation experiments

  20. The neutron radiography facility at Tehran Research Reactor (TRR)

    International Nuclear Information System (INIS)

    Ali Pazirandeh

    2009-01-01

    Full text: Non-destructive testing in many fields of industry including detection of explosives, at the airports, testing for micro-cracks on airplane wings and turbine blades cracks is badly needed. Thermal neutron beam is one of preferable method to detect the micro-cracks, reveals the internal structure of components and explosives. The purpose of this paper is to present the neutron radiography facility at Tehran Research Reactor (TRR), Science and Technology Research Institute, and in particular to emphasize the industrial applications in wood industry, automobile engine inspection, minerals composition identification, turbine blade cracks detection. (author)

  1. Nuclear Security Management for Research Reactors and Related Facilities

    International Nuclear Information System (INIS)

    2016-03-01

    This publication provides a single source guidance to assist those responsible for the implementation of nuclear security measures at research reactors and associated facilities in developing and maintaining an effective and comprehensive programme covering all aspects of nuclear security on the site. It is based on national experience and practices as well as on publications in the field of nuclear management and security. The scope includes security operations, security processes, and security forces and their relationship with the State’s nuclear security regime. The guidance is provided for consideration by States, competent authorities and operators

  2. Pilot-scale reactor activation facility at SRL

    International Nuclear Information System (INIS)

    Bowman, W.W.

    1976-01-01

    The Hydrogeocemical and Stream Sediment Reconnaissance portion of the National Uranium Resource Evaluation program requires an analytical technique for uranium and other elements. Based on an automated absolute activation analysis technique using 252 Cf, a pilt-scale facility installed in a production reactor has provided analyses for 2800 samples. Key features include: an automated sample transport system, a delayed neutron detector, two GeLi detectors, a loader, and an unloader, with all components controlled by a microprocessor; a dedicated PDP-9 computer and pulse height analyzer; and correlation and reduction of acquired data by a series of programs using an IBM 360/195 computer. The facility was calibrated with elemental and isotopic standards. Results of analyses of standard reference materials and operational detection limits for typical sediment samples are presented. Plans to increase sample throughput are discussed briefly

  3. Remote maintenance in TOR fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    Eymery, R.; Constant, M.; Malterre, G.

    1986-11-01

    The TOR facility which is undergoing commissioning tests has a capacity of 5 T. HM/year which is enough for reprocessing all the Phenix fuel, with an excess capacity which is to be used for other fast reactors fuels. It is the result of enlargement and renovation of the old Marcoule pilot facility. A good load factor is expected through the use of equipment with increased reliability and easy maintenance. TOR will also be used to test new equipment developed for the large breeder fuel reprocessing plant presently in the design stage. The latter objective is specifically important for the parts of the plant involving mechanical equipment which are located in a new building: TOR 1. High reliability and flexibility will be obtained in this building thanks to the attention given to the integrated remote handling system [fr

  4. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  5. Annual progress report for 1982 of Theoretical Reactor Physics Section

    International Nuclear Information System (INIS)

    Rastogi, B.P.; Kumar, Vinod

    1983-01-01

    The progress of work done in the Theoretical Reactor Physics Section of the Bhabha Atomic Research Centre, Bombay, during the calendar year 1982 is reported in the form of write-ups and summaries. The main thrust of the work has been to master the neutronic design technology of four different types of nuclear reactor types, namely, pressurized heavy water reactors, boiling light water reactors, pressurized light water reactors and fast breeder reactors. The development work for the neutronic analysis, fuel design, and fuel management of the BWR type reactors of the Tarapur Atomic Power Station has been completed. A new reactor simulator system for PHWR design analysis and core follow-up was completed. Three dimensional static analysis codes based on nodal and finite element methods for the design work of larger size (500-750 MWe) reactors have been developed. Space link kinetics codes in one, two and three dimensions for above-mentioned reactor systems have been written and validated. Fast reactor core disruptive analysis codes have been developed. In the course of R and D work concerning various types of reactor projects, investigations were also carried in the allied areas of Monte Carlo techniques, integral transform methods, path integral methods, high spin states in heavy nuclei and a hydrodynamics model for a laser driven fusion system. (M.G.B.)

  6. 242-A Evaporator crystallizer facility integrated annual safety appraisal

    International Nuclear Information System (INIS)

    1991-01-01

    This report provides the results of the Fiscal Year (FY) 1991 Annual Integrated Safety Appraisal of the 242-A Evaporator Crystallizer Facility in the Hanford 200 East Area. The appraisal was conducted in December 1990 and January 1991, by the Waste Tank Safety Assurance (WTSA) organizations in conjunction with Radiological Engineering, Criticality Safety, Packaging and Shipping Safety, Emergency Preparedness, Environmental Compliance, and Quality Assurance. Reports of these eight organizations are presented as Sections 2 through 7 of this report. The purpose of the appraisal was to verify that the 242-A Evaporator meets US Department of Energy (DOE) and Westinghouse Hanford Company (WHC) requirements and current industry standards of good practice for the areas being appraised. A further purpose was to identify areas in which program effectiveness could be improved. In accordance with the guidance of WHC Management Requirements and Procedures (MRP)5.6, previously identified deficiencies which are being resolved by line management were not repeated as Findings or Observations unless progress or intended disposition was considered to be unsatisfactory

  7. Project RA Research nuclear reactor - Annual report 1993 with comparative review for the period 1991 - 1993

    International Nuclear Information System (INIS)

    1993-12-01

    Research reactor RA Annual report for year 1993 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. The ninth separate annex deals with the feasibility of RA reactor applications. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data [sr

  8. Refueling the RPI reactor facility with low-enrichment fuel

    International Nuclear Information System (INIS)

    Harris, D.R.; Rodriguez-Vera, F.; Wicks, F.E.

    1985-01-01

    The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors. (author)

  9. Status of national programmes on fast reactors 1998/99. 32nd annual meeting. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    Over the past 32 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1998/1999, as reported at the 32. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States.

  10. Status of national programmes on fast reactors 1998/99. 32nd annual meeting. Working material

    International Nuclear Information System (INIS)

    1999-01-01

    Over the past 32 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1998/1999, as reported at the 32. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

  11. 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

  12. 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

  13. 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

  14. On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

    International Nuclear Information System (INIS)

    1989-01-01

    The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  15. The selection of probabilistic safety assessment techniques for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    Vail, J.

    1992-01-01

    Historically, the probabilistic safety assessment (PSA) methodology of choice is the well known event tree/fault tree inductive technique. For reactor facilities is has stood the test of time. Some non-reactor nuclear facilities have found inductive methodologies difficult to apply. The stand-alone fault tree deductive technique has been used effectively to analyze risk in nuclear chemical processing facilities and waste handling facilities. The selection between the two choices suggest benefits from use of the deductive method for non-reactor facilities

  16. The 'SILOE' reactor at Grenoble, France and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the SILOE reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  17. The DR 3 reactor at Risoe, Denmark and its associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DR 2 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of seven information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  18. The DIDO-reactor at Harwell, U.K. and ancillary hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the DIDO reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  19. The 'OSIRIS' reactor at Saclay, France and available hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the OSIRIS reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  20. SANS facility at the Pitesti 14 MW Triga reactor

    International Nuclear Information System (INIS)

    Ionita, I.; Anghel, E.; Mincu, M.; Datcu, A.; Grabcev, B.; Todireanu, S.; Constantin, F.; Shvetsov, V.; Popescu, G.

    2006-01-01

    Full text of publication follows: At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nano-crystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. A monochromatic neutron beam with 1.5 Angstrom ≤ λ ≤ 5 Angstrom is produced by a mechanical velocity selector with helical slots.The distance between sample and detectors plane is (5.2 m ). The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Q min = 0.005 Angstrom -1 while the maximal value is Q max = 0.5 Angstrom -1 . The relative error is ΔQ/Q min = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia. The first step in this cooperation

  1. Annual report on the state of RB reactor components and equipment, december 1999

    International Nuclear Information System (INIS)

    Milosevic, M.

    1999-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 13 times, meaning that experiments were done with 13 configurations of the reactor core. Total reactor operation amounted to 84 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1999, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  2. Annual report on the state of RB reactor components and equipment, december 1998

    International Nuclear Information System (INIS)

    Milosevic, M.

    1998-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  3. Operation and maintenance of the RB reactor, Annual report for 1976

    International Nuclear Information System (INIS)

    Sotic, O.; Vranic, S.

    1976-01-01

    Due to its flexibility and relatively simple construction the RB reactor enabled direct measurements of a series of physical parameters. During 1976 the reactor operation exceeded the plan due to preparation of special experiments planned for the next period. It is planned to operate the reactor at higher power levels (50 W - 10 kw). A need for increasing the neutron flux a neutron converter was built in 1976. preliminary measurements showed that placing the neutron converter next to the reactor vessel enables achievement of irradiation and dosimetry measurements in the fast neutron flux. It is planned to purchase highly enriched fuel for the neutron converter. This annual report includes 5 Annexes with data concerning: operation, irradiation field around the RB reactor, maintenance of reactor components and instrumentation, purchase of new equipment, and the program for training reactor operators

  4. Naval Reactors Facility environmental monitoring report, calendar year 2001

    International Nuclear Information System (INIS)

    2002-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 2001 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U. S. Environmental Protection Agency and the U. S. Department of Energy

  5. 1997 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    1997-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  6. Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003

    International Nuclear Information System (INIS)

    2003-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 2003 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency and the U.S. Department of Energy

  7. Naval Reactors Facility environmental monitoring report, calendar year 1999

    International Nuclear Information System (INIS)

    2000-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1999 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE)

  8. 1993 environmental monitoring report for the naval reactors facility

    International Nuclear Information System (INIS)

    1994-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 1993 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE)

  9. FFTF [Fast Flux Test Facility] reactor shutdown system reliability reevaluation

    International Nuclear Information System (INIS)

    Pierce, B.F.

    1986-07-01

    The reliability analysis of the Fast Flux Test Facility reactor shutdown system was reevaluated. Failure information based on five years of plant operating experience was used to verify original reliability numbers or to establish new ones. Also, system modifications made subsequent to performance of the original analysis were incorporated into the reevaluation. Reliability calculations and sensitivity analyses were performed using a commercially available spreadsheet on a personal computer. The spreadsheet was configured so that future failures could be tracked and compared with expected failures. A number of recommendations resulted from the reevaluation including both increased and decreased surveillance intervals. All recommendations were based on meeting or exceeding existing reliability goals. Considerable cost savings will be incurred upon implementation of the recommendations

  10. Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

    International Nuclear Information System (INIS)

    Haley, T.C.

    2011-01-01

    Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

  11. Reactor cover gas monitoring at the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Bechtold, R.A.; Holt, F.E.; Meadows, G.E.; Schenter, R.E.

    1986-09-01

    The Fast Flux Test Facility (FFTF) is a 400-megawatt (thermal) sodium-cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the US Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100-day operating cycle began in April 1982 and the eighth operating cycle was completed in July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification

  12. A new materials irradiation facility at the Kyoto university reactor

    International Nuclear Information System (INIS)

    Yoshiie, T.; Hayashi, Y.; Yanagita, S.; Xu, Q.; Satoh, Y.; Tsujimoto, H.; Kozuka, T.; Kamae, K.; Mishima, K.; Shiroya, S.; Kobayashi, K.; Utsuro, M.; Fujita, Y.

    2003-01-01

    A new materials irradiation facility with improved control capabilities has been installed at the Kyoto University Reactor (KUR). Several deficiencies of conventional fission neutron material irradiation systems have been corrected. The specimen temperature is controlled both by an electric heater and by the helium pressure in the irradiation tube without exposure to neutrons at temperatures different from the design test conditions. The neutron spectrum is varied by the irradiation position. Irradiation dose is changed by pulling the irradiation capsule up and down during irradiation. Several characteristics of the irradiation field were measured. The typical irradiation intensity is 9.4x10 12 n/cm 2 s (>0.1 MeV) and the irradiation temperature of specimens is controllable from 363 to 773 K with a precision of ±2 K

  13. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    International Nuclear Information System (INIS)

    Lewis, M.R.

    2000-01-01

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork

  14. 1991 environmental monitoring report for the Naval Reactors Facility

    International Nuclear Information System (INIS)

    1991-01-01

    The results of the radiological and non-radiological environmental monitoring programs for 1991 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were within the guidelines established by state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or heath and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the EnVironmental Protection Agency (EPA) and the Department of Energy (DOE)

  15. Naval Reactors Facility environmental monitoring report, calendar year 2000

    International Nuclear Information System (INIS)

    2001-01-01

    The results of the radiological and nonradiological environmental monitoring programs for 2000 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE)

  16. International Working Group on Fast Reactors Second Annual Meeting. Summary Report

    International Nuclear Information System (INIS)

    1969-01-01

    The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

  17. Annual Report of Radioactive Waste Facilities Operation in 2015

    Institute of Scientific and Technical Information of China (English)

    DU; Hong-ming; GAO; Zhi-gang; DIAO; Lei; SHEN; Zheng; LI; Wen-ge

    2015-01-01

    301of the Department of Radiochemistry,is in charge of the management of radioactive waste and the safety of the relative facilities to meet the request of the scientific research production.There are 16radioactive waste facilities,including9facilities which are closed and monitored

  18. Reactor cover gas monitoring at the Fast Flux Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bechtold, R A; Holt, F E; Meadows, G E; Schenter, R E [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    The Fast Flux Test Facility (FFTF) is a 400 megawatt (thermal) sodium cooled reactor designed for irradiation testing of fuels, materials and components for LMRs. It is operated by the Westinghouse Hanford Company for the U. S. Department of Energy on the government-owned Hanford reservation near Richland, Washington. The first 100 day operating cycle began in April 1982 and the eighth operating cycle was completed In July 1986. Argon is used as the cover gas for all sodium systems at the plant. A program for cover gas monitoring has been in effect since the start of sodium fill in 1978. The argon is supplied to the FFTF by a liquid argon Dewar System and used without further purification. A liquid argon Dewar system provides the large volume of inert gas required for operation of the FFTF. The gas is used as received and is not recycled. Low concentrations of krypton and xenon in the argon supply are essential to preclude interference with the gas tag system. Gas chromatography has been valuable for detection of inadvertent air in leakage during refueling operations. A temporary system is installed over the reactor during outages to prevent oxide formation in the sodium vapor traps upstream from the on line gas chromatograph. On line gas monitoring by gamma spectrometry and grab sampling with GTSTs has been successful for the identification of numerous radioactive gas releases from creep capsule experiments as well as 9 fuel pin ruptures. A redundant fission gas monitoring system has been installed to insure constant surveillance of the reactor cover gas.

  19. Program for studying fundamental interactions at the PIK reactor facilities

    International Nuclear Information System (INIS)

    Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.

    2016-01-01

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  20. Program for studying fundamental interactions at the PIK reactor facilities

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Vassiljev, A. V.; Varlamov, V. E. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Geltenbort, P. [Institut Laue-Langevin (France); Gridnev, K. A. [St. Petersburg State University (Russian Federation); Dmitriev, S. P.; Dovator, N. A. [Russian Academy of Sciences, Ioffe Physical-Technical Institute (Russian Federation); Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A. [National Research Center Kurchatov Institute, Petersburg Nuclear Physics Institute (Russian Federation); Martemyanov, V. P. [National Research Center Kurchatov Institute (Russian Federation); and others

    2016-05-15

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4′ channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4′ channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  1. On exposure management of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1993

    International Nuclear Information System (INIS)

    1994-01-01

    The Law of Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1993 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the the exposure of workers was below the permissible exposure dose in 1993 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of JAERI and PNC. (J.P.N.)

  2. University of Wisconsin, Nuclear Reactor Laboratory. Annual report, 1985-1986

    International Nuclear Information System (INIS)

    Cashwell, R.J.

    1986-01-01

    Operational activities for the reactor are described concerning nuclear engineering classes from the University of Wisconsin; reactor sharing program; utility personnel training; sample irradiations and neutron activation analysis; and changes in personnel, facility, and procedures. Results of surveillance tests are presented for operating statistics and fuel exposure; emergency shutdowns and inadvertent scrams; maintenance; radioactive waste disposal; radiation exposures; environmental surveys; and publications and presentations on work based on reactor use

  3. Status of safety at Areva group facilities. 2006 annual report

    International Nuclear Information System (INIS)

    2006-01-01

    This report presents a snapshot of nuclear safety and radiation protection conditions in the AREVA group's nuclear installations in France and abroad, as well as of radiation protection aspects in service activities, as identified over the course of the annual inspections and analyses program carried out by the General Inspectorate in 2006. This report is presented to the AREVA Supervisory Board, communicated to the labor representation bodies concerned, and made public. In light of the inspections, appraisals and coordination missions it has performed, the General Inspectorate considers that the nuclear safety level of the AREVA group's nuclear installations is satisfactory. It particularly noted positive changes on numerous sites and efforts in the field of continuous improvement that have helped to strengthen nuclear safety. This has been possible through the full involvement of management teams, an improvement effort initiated by upper management, actions to increase personnel awareness of nuclear safety culture, and supervisors' heightened presence around operators. However, the occurrence of certain events in facilities has led us to question the nuclear safety repercussions that the changes to activities or organization on some sites have had. In these times of change, drifts in nuclear safety culture have been identified. The General Inspectorate considers that a preliminary analysis of the human and organizational factors of these changes, sized to match the impact the change has on nuclear safety, should be made to ensure that a guaranteed level of nuclear safety is maintained (allowance for changes to references, availability of the necessary skills, resources of the operating and support structures, etc.). Preparations should also be made to monitor the changes and spot any telltale signs of drift in the application phase. Managers should be extra vigilant and the occurrence of any drift should be systematically dealt with ahead of implementing

  4. Reactor safety. Annual technical progress report, Government fiscal year 1979

    International Nuclear Information System (INIS)

    1980-01-01

    Information is presented on LMFBR reactor safety concerning the energetics effects of sodium spray fires; sodium drop and spray burning; core debris accommodation; attenuation in containment; and attenuation in the environment

  5. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author).

  6. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    International Nuclear Information System (INIS)

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  7. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    International Nuclear Information System (INIS)

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department's programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  8. Reactor engineering department annual report. April 1, 1994 - March 31, 1995

    International Nuclear Information System (INIS)

    1995-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1994 (April 1, 1994 - March 31, 1995). The major Department's programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basic researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, fusion neutronics, radiation shielding, reactor instrumentation, reactor control/diagnosis, thermohydraulics and technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal-hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  9. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    Energy Technology Data Exchange (ETDEWEB)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department`s programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  10. Reactor Engineering Department annual report (April 1, 1996 - March 31, 1997)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-10-01

    This report summarizes the research and development activities in the Reactor Engineering Department of JAERI during the fiscal year of 1996 (April 1, 1996 - March 31, 1997). The major Department`s programs promoted in the year are the design activities of advanced reactor system and the development of a high power proton linear accelerator to construct an intense neutron source for innovative neutron science. Other Major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analysis, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC`s fast reactor project were also progressed. The 99 papers are indexed individually. (J.P.N.)

  11. Reactor Engineering Department annual report. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    Ochiai, Masaaki; Ohnuki, Akira; Ono, Toshihiko

    1998-11-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1997 (April 1, 1997 - March 31, 1998). The major Department's programs promoted in the year are the achievement of the world-strongest lasing of Free Electron Laser and the verification of the core thermal integrity during design basis events in PWRs. Other Major tasks of the Department are various basic researches on the advanced reactor system design studies, the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the reactor instrumentation, the reactor control/diagnosis, the thermal hydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermal hydraulic facilities. The cooperative works to JAERI's major projects such as the high temperature gas cooled reactor, the fusion reactor and PNC's fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  12. Reactor engineering department annual report. April 1, 1995 - March 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1995 (April 1, 1995 - March 31, 1996). The major Department`s programs promoted in the year are the design activities of advanced reactor system and development of a high intensity proton linear accelerator for the engineering applications including TRU incineration. Other major tasks of the Department are various basics researches on the nuclear data and group constants, the developments of theoretical methods and codes, the reactor physics experiments and their analyses, the fusion neutronics, the radiation shielding, the reactor instrumentation, the reactor control/diagnosis, the thermalhydraulics and the technology developments related to the reactor engineering facilities, the accelerator facilities and the thermalhydraulic facilities. The cooperative works to JAERI`s major projects such as the high temperature gas cooled reactor or the fusion reactor and to PNC`s fast reactor project were also progressed. The activities of the research committees to which the Department takes a role of secretariat are also summarized in this report. (author)

  13. 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

  14. Upgrade of the experimental facilities of the ORPHEE reactor

    International Nuclear Information System (INIS)

    Farnoux, B.; Breant, P.

    1993-01-01

    At the time of the design, the ORPHEE reactor has been equipped with a set of up-to-date experimental facilities such as nine tangential and horizontal beam holes, one hot source, two hydrogen cold sources and six neutron guides. After more than ten years of operations, all the neutron beams are now used by about twenty five spectrometers. A modernisation program is under progress with a two fold aim: upgrade of the existing facilities and creation of new beams. Some details of the six following points will be described: 1) replacement of the flat cold source cell by an hollow cylinder in order first to increase the cold neutron flux and secondly to facilitate the extraction of new cold neutron beams. 2) replacement of the old neutron guide elements coated with natural nickel by new elements with isotopic nickel or super mirror coating. 3) modification of the curvature of some existing neutron guides in order to increase the wavelength band transmission. 4) creation of new cold neutron beams by installation of benders on the existing neutron guides. 5) design of new cold neutron guides and a new guide hall. 6) design of a thermal neutron guide. The two last points will made extensive use of super mirrors allowed by new technical developments done at the Laboratoire LEON BRILLOUIN in connection with industry. (author)

  15. State of the art of nuclear facilities with organic cooled reactors

    International Nuclear Information System (INIS)

    Brede, O.

    1984-01-01

    USA, Canadian, and USSR activities aimed at developing nuclear facilities with organic cooled reactors are summarized. The facilities OMRE, PNPF, WR-1, and ARBUS are described, discussing in particular the problems of the chemistry of organic coolants. Finally, problems of further development and prospects of the application of organic cooled reactors are briefly outlined. (author)

  16. Analysis of Opportunity to Create Self-Regulating Reactor Facility of Extra-Low Power

    International Nuclear Information System (INIS)

    Kazansky, Y.A.; Levtchenko, V.A.; Yuriev, Y.S.

    2002-01-01

    This paper deals with fundamental possibilities (economy, safety, self-regulation) of creating an extra-low power reactor facility for heat supply. It contains the results of calculations for thermal and fast neutron reactors. The concept of this type of a reactor had been developed by the contributors earlier

  17. 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

  18. Conceptual design of a mirror reactor for a fusion engineering research facility (FERF)

    International Nuclear Information System (INIS)

    Batzer, T.H.; Burleigh, R.C.; Carlson, G.A.; Dexter, W.L.; Hamilton, G.W.; Harvey, A.R.; Hickman, R.G.; Hoffman, M.A.; Hooper, E.B. Jr.; Moir, R.W.; Nelson, R.L.; Pittenger, L.C.; Smith, B.H.; Taylor, C.E.; Werner, R.W.; Wilcox, T.P.

    1975-01-01

    A conceptual design is presented for a small mirror fusion reactor for a Fusion Engineering Research Facility (FERF). The reactor produces 3.4 MW of fusion power and a useful neutron flux of about 10 14 n.cm -2 .s -1 . Superconducting ''yin-yang'' coils are used, and the plasma is sustained by injection of energetic neutral D 0 and T 0 . Conceptual layouts are given for the reactor, its major components, and supporting facilities. (author)

  19. Annual report to the Laser Facility Committee, 1982

    International Nuclear Information System (INIS)

    1982-03-01

    The report covers the work done at, or in association with, the Central Laser Facility during the year April 1981 to March 1982 under the headings; glass laser facility development, gas laser development, laser plasma interactions, transport and particle emission studies, ablative acceleration and compression studies, spectroscopy and XUV lasers, and, theory and computation. Publications based on the work of the facility which have either appeared or been accepted for publication during the year are listed. (U.K.)

  20. The BNCT facility at the HFR Petten: Quality assurance for reactor facilities in clinical trials

    International Nuclear Information System (INIS)

    Moss, R.; Watkins, P.; Vroegindeweij, C.; Stecher-Rasmussen, F.; Huiskamp, R.; Ravensberg, K.; Appelman, K.; Sauerwein, W.; Hideghety, K.; Gabel, D.

    2001-01-01

    The first clinical trial in Europe of Boron Neutron Capture Therapy (BNCT) for the treatment of glioblastoma was opened in July 1997. The trial is a Phase I study with the principal aim to establish the maximum tolerated radiation dose and the dose limiting toxicity under defined conditions. It is the first time that a clinical application could be realised on a completely multi-national scale. The treatment takes place at the High Flux Reactor (HFR) in Petten, the Netherlands, is operated by an international team of experts under the leadership of a German radiotherapist, and treats patients coming from different European countries. It has therefore been necessary to create a very specialised organisation and contractual structure with the support of administrations from different countries, who had to find and adapt solutions within existing laws that had never foreseen such a situation. Furthermore, the treatment does not take place in an hospital environment and even more so, the facility is at a nuclear research reactor. Hence, special efforts were made on quality assurance, in order that the set-up at the facility and the personnel involved complied, as closely as possible, with similar practices in conventional radiotherapy departments. (author)

  1. An independent safety assessment of Department of Energy nuclear reactor facilities: Safety overview and management function

    International Nuclear Information System (INIS)

    Booth, M.; Brodsky, R.S.; Frankhouser, W.L.

    1981-02-01

    The Under Secretary of Energy established the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee in October, 1979, in the aftermath of the Three Mile Island (TMI) nuclear accident, to assess the adequacy of training of personnel at DOE nuclear facilities. Subsequently, in February, 1980, the charge to this Committee was modified to assess all implications of the Kemeny Commission report on TMI with regard to DOE nuclear reactors, excluding those in the Division of Naval Reactors. The modified charge was also limited, for the time being, to reactor facilities instead of all nuclear facilities. This report describes the portion of the revised assessment activities that was assigned to the Assessment Support Team

  2. Department of Nuclear Safety Research and Nuclear Facilities annual report 1995

    Energy Technology Data Exchange (ETDEWEB)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Jacobsen, U.; Oelgaard, P.L. [eds.

    1996-03-01

    The report presents a summary of the work of the Department of Nuclear Safety Research and Nuclear Facilities in 1995. The department`s research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff`s participation in national and international committees. (au) 5 tabs., 21 ills.

  3. Department of Nuclear Safety Research and Nuclear Facilities annual report 1995

    International Nuclear Information System (INIS)

    Majborn, B.; Brodersen, K.; Damkjaer, A.; Floto, H.; Jacobsen, U.; Oelgaard, P.L.

    1996-03-01

    The report presents a summary of the work of the Department of Nuclear Safety Research and Nuclear Facilities in 1995. The department's research and development activities are organized in three research programmes: Radiation Protection, Reactor Safety, and Radioanalytical Chemistry. The nuclear facilities operated by the department include the Research Reactor DR3, the Isotope Laboratory, the Waste Treatment Plant, and the Educational Reactor DR1. Lists of staff and publications are included together with a summary of the staff's participation in national and international committees. (au) 5 tabs., 21 ills

  4. Concerning control of radiation exposure to workers in nuclear reactor facilities for testing and nuclear reactor facilities in research and development phase (fiscal 1987)

    International Nuclear Information System (INIS)

    1988-01-01

    A nuclear reactor operator is required by the Nuclear Reactor Control Law to ensure that the radiation dose to workers engaged in the operations of his nuclear reactor is controlled below the permissible exposure doses that are specified in notifications issued based on the Law. The present note briefly summarizes the data given in the Reports on Radiation Control, which have been submitted according to the Nuclear Reactor Control Law by the operators of nuclear reactor facilities for testing and those in the research and development phase, and the Reports on Control of Radiation Exposure to Workers submitted in accordance with the applicable administrative notices. According to these reports, the measured exposure to workers in 1987 were below the above-mentioned permissible exposure doses in all these nuclear facilities. The 1986 and 1987 measurements of radiation exposure dose to workers in nuclear reactor facilities for testing are tabulated. The measurements cover dose distribution among the facilities' personnel and workers of contractors. They also cover the total exposure dose for all workers in each of four plants operated under the Japan Atomic Energy Research Institute and the Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

  5. Annual report on operation and management of hot laboratories and facilities. From April 1, 2006 to March 31, 2007

    International Nuclear Information System (INIS)

    2008-02-01

    This is an annual report in a fiscal year 2006 that describes activities of the Reactor Fuel Examination Facility (RFEF), the Waste Safety Testing Facility (WASTEF), the Research Hot Laboratory (RHL) and the other research hot facilities in the Department of Hot laboratories and facilities. In RFEF, destructive examinations of BWR fuel rods and re-assembly were carried out as PIEs for a fuel assembly irradiated for 5 cycles in the Fukushima-2 Nuclear Power Station Unit-1. Mechanical property measurement of high burn-up fuel rods were performed as spent fuel integrity test for long term dry storage in order to formulate guidelines and technical criteria. In WASTEF, Slow Strain Rate Tests (SSRT) and Uni-axial Constant Load Tensile tests (UCLT) of in-core materials in pressurized high-temperature water condition, stress corrosion cracking tests for high-performance fuel cladding material and calorific value measurement of pulse irradiated fuel in NSRR were carried out. In RHL, equipment un-installations and decontamination were performed to lead cells according to the decommissioning plan. And modification of fuel storage room were started in order to utilize the facility for un-irradiated fuel storage after a fiscal year 2007. In addition, management of the other research hot facilities (No.1 Plutonium Laboratory, No.2 Research Laboratory, No.4 Research Laboratory, Analytical Chemistry Laboratory, Uranium Enrichment Laboratory, (Simulation Test for Environmental Radionuclide Migration (STEM), Clean Laboratory for Environmental Analysis and Research (CLEAR) and fuel storage) were carried out. (author)

  6. 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

  7. Description of the RA-3 research reactor as a model facility

    International Nuclear Information System (INIS)

    Vicens, Hugo E.; Quintana, Jorge A.

    2001-01-01

    The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

  8. Annual harvests of Corbicula populations prevent clogging of nuclear reactor heat exchangers

    International Nuclear Information System (INIS)

    Harvey, R.S.

    1983-01-01

    An annual program for removal of millions of Corbicula from upstream cooling water basins has prevented reclogging of nuclear reactor heat exchanger distributor plates at the Savannah River Plant during the past seven years. There are nine 32-megaliter basins in the three operating reactor areas where some settling of particulates occurs before cooling water is passed through screens in route to heat exchangers. Annual cleanings keep silt/clam substrate levels low and clam sizes small. Data are presented on the size/age distribution for clams recolonizing basins between cleanings

  9. 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

  10. Integral Fast Reactor Program annual progress report, FY 1991

    International Nuclear Information System (INIS)

    1992-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1991. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

  11. Integral Fast Reactor Program. Annual progress report, FY 1992

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1993-06-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R&D.

  12. Fusion reactors: physics and technology. Annual progress report

    International Nuclear Information System (INIS)

    Conn, R.W.

    1983-08-01

    Fusion reactors are designed to operate at full power and generally at steady state. Yet experience shows the load variations, licensing constraints, and frequent sub-system failures often require a plant to operate at fractions of rated power. The aim of this study has been to assess the technology problems and design implications of startup and fractional power operation on fusion reactors. The focus of attention has been tandem mirror reactors (TMR) and we have concentrated on the plasma and blanket engineering for startup and fractional power operation. In this report, we first discuss overall problems of startup, shutdown and staged power operation and their influence on TMR design. We then present a detailed discussion of the plasma physics associated with TMR startup and various means of achieving staged power operation. We then turn to the issue of instrumentation and safety controls for fusion reactors. Finally we discuss the limits on transient power variations during startup and shutdown of Li 17 Pb 83 cooled blankets

  13. Integral Fast Reactor Program annual progress report, FY 1994

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, J.J.

    1994-12-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1994. Technical accomplishments are presented in the following areas of the IFR technology development activities: metal fuel performance; pyroprocess development; safety experiments and analyses; core design development; fuel cycle demonstration; and LMR technology R ampersand D

  14. Annual report 1980 of the Research and Measurement Reactor Braunschweig

    International Nuclear Information System (INIS)

    Kriks, H.J.; Vorbrugg, W.

    1981-08-01

    In 1980 the Forschungs- und Messreaktor Braunschweig (FMRB) has been in action for 2457 hours without any serious disturbance. The experience in the field of reactor operation, in the radiation protection work as well as in the experiments at the beam tubes is reported. (orig.) [de

  15. Integral Fast Reactor Program. Annual progress report, FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D.

  16. Integral Fast Reactor Program. Annual progress report, FY 1993

    International Nuclear Information System (INIS)

    Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

    1994-10-01

    This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1993. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R and D

  17. Participation in the US Department of Energy Reactor Sharing Program. Annual report, September 30, 1993--September 29, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.

  18. Dual Axis Radiographic Hydrodynamic Test Facility mitigation action plan. Annual report for 1998

    Energy Technology Data Exchange (ETDEWEB)

    Haagenstad, T.

    1999-01-15

    This Mitigation Action Plan Annual Report (MAPAR) has been prepared as part of implementing the Dual Axis Radiographic Hydrodynamic Test Facility (DARHT) Mitigation Action Plan (MAP) to protect workers, soils, water, and biotic and cultural resources in and around the facility.

  19. Dual Axis Radiographic Hydrodynamic Test Facility mitigation action plan. Annual report for 1998

    International Nuclear Information System (INIS)

    Haagenstad, T.

    1999-01-01

    This Mitigation Action Plan Annual Report (MAPAR) has been prepared as part of implementing the Dual Axis Radiographic Hydrodynamic Test Facility (DARHT) Mitigation Action Plan (MAP) to protect workers, soils, water, and biotic and cultural resources in and around the facility

  20. Annual Report of Radioactive Waste Facilities Operation in 2013

    Institute of Scientific and Technical Information of China (English)

    DU; Hong-ming; GAO; Zhi-gang; LIU; Fu-guo

    2013-01-01

    301,a section of Department of Radiochemistry,which manages 15 facilities and undertakes the administrative tasks of radioactive waste,is the important guarantee of scientific research production and safety in CIAE.1 The safe operation of the radioactive waste management facilities In 2013,in order to ensure the operation safety,we formulated the inspection regulations,which included regular operation inspection,week safety inspection from the leaders of the section and

  1. ATR National Scientific User Facility 2009 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Todd R. Allen; Mitchell K. Meyer; Frances Marshall; Mary Catherine Thelen; Jeff Benson

    2010-11-01

    This report describes activities of the ATR NSUF from FY-2008 through FY-2009 and includes information on partner facilities, calls for proposals, users week and education programs. The report also contains project information on university research projects that were awarded by ATR NSUF in the fiscal years 2008 & 2009. This research is university-proposed researcher under a user facility agreement. All intellectual property from these experiments belongs to the university per the user agreement.

  2. Annual report to the Laser Facility Committee 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The report covers the work done at, or in association with, the Central Laser Facility during the year ended 31 March 1983. There are eight chapters in all, six corresponding to the six groups of the Glass Laser Scientific Programme and Scheduling Committee, a chapter on gas laser development, and a chapter describing the work and development of the newly established Ultraviolet Radiation Facility. (author)

  3. ATR National Scientific User Facility 2009 Annual Report

    International Nuclear Information System (INIS)

    Allen, Todd R.; Meyer, Mitchell K.; Marshall, Frances; Thelen, Mary Catherine; Benson, Jeff

    2010-01-01

    This report describes activities of the ATR NSUF from FY-2008 through FY-2009 and includes information on partner facilities, calls for proposals, users week and education programs. The report also contains project information on university research projects that were awarded by ATR NSUF in the fiscal years 2008 and 2009. This research is university-proposed researcher under a user facility agreement. All intellectual property from these experiments belongs to the university per the user agreement.

  4. Federal Facility Agreement Annual Progress Report for FY 1998

    Energy Technology Data Exchange (ETDEWEB)

    Palmer, E.

    1999-08-04

    This FFA Annual Progress Report has been developed to summarize the information for activities performed during the Fiscal Year 1998 (October 1, 1997, to September 30, 1998) and activities planned for Fiscal Year 1999 by U.S. EPA, SCDHEC, and SRS at those units and areas identified for remediation in the Agreement.

  5. Federal Facility Agreement Annual Progress Report for Fiscal Year 1998

    International Nuclear Information System (INIS)

    Palmer, E.

    1999-01-01

    This FFA Annual Progress Report has been developed to summarize the information for activities performed during the Fiscal Year 1998 (October 1, 1997, to September 30, 1998) and activities planned for Fiscal Year 1999 by U.S. EPA, SCDHEC, and SRS at those units and areas identified for remediation in the Agreement

  6. Fermilab Testbeam Facility Annual Report – FY 2016

    Energy Technology Data Exchange (ETDEWEB)

    Albrow, M. G. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States). et al.

    2016-11-01

    This Technical Memorandum (TM) summarizes the Fermilab Test Beam operations for FY 2015. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the individual experiments that ran at FTBF, which are tabulated. Each experiment section was prepared by the relevant authors, and was edited for inclusion in this summary.

  7. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  8. Seismic response analyses for reactor facilities at Savannah River

    International Nuclear Information System (INIS)

    Miller, C.A.; Costantino, C.J.; Xu, J.

    1991-01-01

    The reactor facilities at the Savannah River Plant (SRP) were designed during the 1950's. The original seismic criteria defining the input ground motion was 0.1 G with UBC [uniform building code] provisions used to evaluate structural seismic loads. Later ground motion criteria have defined the free field seismic motion with a 0.2 G ZPA [free field acceleration] and various spectral shapes. The spectral shapes have included the Housner spectra, a site specific spectra, and the US NRC [Nuclear Regulatory Commission] Reg. Guide 1.60 shape. The development of these free field seismic criteria are discussed in the paper. The more recent seismic analyses have been of the following type: fixed base response spectra, frequency independent lumped parameter soil/structure interaction (SSI), frequency dependent lumped parameter SSI, and current state of the art analyses using computer codes such as SASSI. The results from these computations consist of structural loads and floor response spectra (used for piping and equipment qualification). These results are compared in the paper and the methods used to validate the results are discussed. 14 refs., 11 figs

  9. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  10. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY: Nuclear Regulatory Commission... Commission (NRC or the Commission) has issued renewed Facility Operating License No. R- 112, held by Reed... License No. R-112 will expire 20 years from its date of issuance. The renewed facility operating license...

  11. Fusion Reactor Safety Research Program annual report, FY-79

    International Nuclear Information System (INIS)

    Crocker, J.G.; Cohen, S.

    1980-08-01

    The objective of the program is the development, coordination, and execution of activities related to magnetic fusion devices and reactors that will: (a) identify and evaluate potential hazards, (b) assess and disclose potential environmental impacts, and (c) develop design standards and criteria that eliminate, mitigate, or reduce those hazards and impacts. The program will provide a sound basis for licensing fusion reactors. Included in this report are portions of four reports from two outside contractors, discussions of the several areas in which EG and G Idaho is conducting research activities, a discussion of proposed program plan development, mention of special tasks, a review of fusion technology program coordination by EG and G with other laboratories, and a brief view of proposed FY-80 activities

  12. The Annual Neutron School: Program and Facility for Nuclear Science and Technology

    International Nuclear Information System (INIS)

    Dingle, C.A.M.; Bautista, U.M.; Jecong, J.F.M.; Hila, F.C.; Astronomo, A.A.; Olivares, R.U.; Guillermo, N.R.D.; Ramo, M.E.S.K.V.; Saligan, P.P.

    2015-01-01

    The core realization of the mandate of the Philippine Nuclear Research Institute (PNRI) is the establishment and utilization of major nuclear facilities in lieu of the decommissioned research reactor. To address the need for manpower in the future, the applied physics research section (APRS) of the PNRI has initiated capacity building in the use and operation of small neutron sources which attempts to re-establish, develop and sustain expertise in nuclear science and technology. These activities have provided the theoretical and experimental training of young professionals and scientist of the institute which, consequently, resulted in the conceptualization of the Annual Neutron School (ANS).The ANS provides training and teaching environments for the young generation who will operate, utilize and regulate future nuclear facilities. More importantly, it demonstrates and presents the acquired knowledge and research outputs by the staff via “train a trainer” concept to an audience of junior undergraduate students. The successful implementation of the ANS has been participated by selected universities within Metro Manila and was able to train a number of students since its establishment in 2013. The program offers training, education, and R & D in the basic nuclear instrumentation and techniques which includes (1) characterization of different neutron sources – AmBe, PuBe and Cf-252; (2) development of Neutron Activation Analysis (NAA) technique using a portable neutron source for non-destructive elemental analysis; (3) utilization of MCNP (Monte Carlo N-Particle) code for verification of experimental data on neutron characterization, radiation dosimetry, detector design, calibration and efficiency and TRIGA fuel assembly configuration for sub-critical experiments. (author)

  13. 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

  14. Annual report on the state of RB reactor components and equipment, December 1997

    International Nuclear Information System (INIS)

    Milosevic, M.

    1997-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1997 the reactor lattice was not changed due to application of the coupled fast-thermal core HERBE. Total reactor operation amounted to 69.5 Wh with 66 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment, plan for forming new HERBE core, plan for regular annual maintenance of the reactor, and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  15. An independent safety assessment of Department of Energy nuclear reactor facilities: Procedures, operations and maintenance

    International Nuclear Information System (INIS)

    Toto, G.; Lindgren, A.J.

    1981-02-01

    The 1979 accident at the Three Mile Island commercial nuclear power plant has led to a number of studies of nuclear reactors, in both the public and private sectors. One of these is that of the Department of Energy's (DOE) Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee, which has outlined tasks for assessment of 13 reactors owned by DOE and operated by contractors. This report covers one of the tasks, the assessment of procedures, operations, and maintenance at the DOE reactor facilities, based on a review of actual documents used at the reactor sites

  16. Radiation protection at the RA Reactor in 1998, RA reactor annual report, Part -2

    International Nuclear Information System (INIS)

    Ninkovic, M.; Pavlovic, R.; Mandic, M.; Pavlovic, S.; Grsic, Z.

    1998-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Collecting and treatment of fluid effluents; and (4) radioactive wastes, decontamination and actions. Each of the category is described as a separate annex of this report [sr

  17. Radiation protection at the RA Reactor in 1988, Part -2, RA reactor annual report

    International Nuclear Information System (INIS)

    Ninkovic, M.; Ajdacic, N.; Zaric, M.; Vukovic, Z.

    1988-01-01

    Radiation protection tasks which enable safe operation of the RA reactor, and are defined according the the legal regulations and IAEA safety recommendations are sorted into four categories in this report: (1) Control of the working environment, dosimetry at the RA reactor and radiation protection; (2) Radioactivity control in the vicinity of the reactor and meteorology measurements; (3) Decontamination and relevant actions, collecting and treatment of fluid effluents; and and solid radioactive wastes [sr

  18. Research nuclear reactor RA - Annual Report 1995, with comparative review for period 1991-1995

    International Nuclear Information System (INIS)

    Sotic, O.

    1995-12-01

    Activities related to revitalisation of the RA reactor started in 1986, were continued during period 1991-1995. All the planned actions related to renewal of the reactor components were finished. The last, and at the same time most important action, related to exchange of complete reactor instrumentation is underway, delayed. The delivery of components from USSR is late. Production of this instruments is financed by the IAEA according to the contract signed in December 1988 with Russian Atomenergoexport. According to this contract, it has been planned that the RA reactor instrumentation should be delivered to the Vinca Institute by the end of 1990. Only 56% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the temporary embargo imposed by the IAEA. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 47 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved during the period 1991-1995. This research reactor RA Annual is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection [sr

  19. Advanced Test Reactor (ATR) Facility 10CFR830 Safety Basis Related to Facility Experiments

    International Nuclear Information System (INIS)

    Tomberlin, T.A.

    2002-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) Advanced Test Reactor (ATR), a DOE Category A reactor, was designed to provide an irradiation test environment for conducting a variety of experiments. The ATR Safety Analysis Report, determined by DOE to meet the requirements of 10 CFR 830, Subpart B, provides versatility in types of experiments that may be conducted. This paper addresses two general types of experiments in the ATR facility and how safety analyses for experiments are related to the ATR safety basis. One type of experiment is more routine and generally represents greater risks; therefore this type of experiment is addressed with more detail in the safety basis. This allows individual safety analyses for these experiments to be more routine and repetitive. The second type of experiment is less defined and is permitted under more general controls. Therefore, individual safety analyses for the second type of experiment tend to be more unique from experiment to experiment. Experiments are also discussed relative to ''major modifications'' and DOE-STD-1027-92. Application of the USQ process to ATR experiments is also discussed

  20. Annual report to the Laser Facility Committee 1984

    International Nuclear Information System (INIS)

    1984-01-01

    The report describes the work carried out at, or in association with, the Central Laser Facility (CLF), during the year ending March 1984. The CLF programme is divided into three main sections. The first, the glass laser scientific programme, is concerned with applications of the high power Nd glass laser. The second, the ultra violet radiation facility scientific programme, involves the excimer pumped frequency tunable lasers. The last, high power KrF laser development, describes Research and development work on this laser. (U.K.)

  1. Annual report to the Laser Facility Committee 1979

    International Nuclear Information System (INIS)

    1979-03-01

    The report covers the work done at the Central Laser Facility, Rutherford Laboratory during the year preceding 31 March 1979. Preliminary work already undertaken on the upgrade of the glass laser and target areas consisting of the relocation of the two beam target chamber and tests on phosphate glass and also the completion of the electron beam generator for use by researchers on high power gas laser systems, are described. Work of the groups using the glass laser facility are considered under the headings; glass laser development, gas laser development, laser plasma interactions, transport and particle emission, ablative compression studies, atomic and radiation physics, XUV lasers, theory and computation. (U.K.)

  2. Fermilab Test Beam Facility Annual Report FY17

    Energy Technology Data Exchange (ETDEWEB)

    Rominsky, M. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Schmidt, E. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Rivera, R. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Uplegger, L. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Asaadi, J. [Univ. of Texas, Arlington, TX (United States); Raaf, J. L. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Freeman, J. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Price, J. [Univ. of Liverpool (United Kingdom); Casey, B. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Ehrlich, R. [Univ. of Virginia, Charlottesville, VA (United States); Belmont, R. [Univ. of Colorado, Boulder, CO (United States); Boose, S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Conners, M. [Georgia State Univ., Atlanta, GA (United States); Haggerty, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hill, K. [Univ. of Colorado, Boulder, CO (United States); Hodges, A. [Georgia State Univ., Atlanta, GA (United States); Huang, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Kistenev, E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Lajoie, J. [Iowa State Univ., Ames, IA (United States); Mannel, E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Osborn, J. [Univ. of Michigan, Ann Arbor, MI (United States); Pontieri, C. [Brookhaven National Lab. (BNL), Upton, NY (United States); Purschke, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Sarsour, M. [Georgia State Univ., Atlanta, GA (United States); Sen, A. [Iowa State Univ., Ames, IA (United States); Skoby, M. [Univ. of Michigan, Ann Arbor, MI (United States); Stoll, S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Toldo, F. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ujvari, B. [Debrecen Univ., Debrecen (Hungary); Woody, C. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ronzhin, A. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Hanagaki, K. [High Energy Accelerator Research Organization (KEK), Tsukuba (Japan); Apresyan, A. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Bose, T. [Boston Univ., MA (United States); Canepa, A. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Demina, R. [Univ. of Rochester, NY (United States); Gershtein, Y. [Rutgers Univ., Piscataway, NJ (United States); Halkiadakis, E. [Rutgers Univ., Piscataway, NJ (United States); Haytmyradov, M. [Univ. of Iowa, Iowa City, IA (United States); Hazen, E. [Boston Univ., MA (United States); Hindrichs, O. [Univ. of Rochester, NY (United States); Korjenevski, S. [Univ. of Rochester, NY (United States); Nachtman, J. [Univ. of Iowa, Iowa City, IA (United States); Narain, M. [Brown Univ., Providence, RI (United States); Nash, K. [Rutgers Univ., Piscataway, NJ (United States); Onel, Y. [Univ. of Iowa, Iowa City, IA (United States); Osherson, M. [Rutgers Univ., Piscataway, NJ (United States); Rankin, D. [Boston Univ., MA (United States); Schneider, B. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Stone, B. [Rutgers Univ., Piscataway, NJ (United States); Metcalfe, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Benoit, M. [Univ. of Geneva (Switzerland); Vicente, M. [Univ. of Geneva (Switzerland); di Bello, F. [Univ. of Geneva (Switzerland); Cavallaro, E. [Univ. Autonoma de Barcelona (Spain); Chakanov, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Frizzell, D. [Univ. of Oklahoma, Norman, OK (United States); Kiehn, M. [Univ. of Geneva (Switzerland); Meng, L. [Univ. of Geneva (Switzerland); Miucci, A. [Univ. of Bern, Bern (Switzerland); Nodulman, L. [Argonne National Lab. (ANL), Argonne, IL (United States); Terzo, S. [Univ. Autonoma de Barcelona (Spain); Wang, Rui [Argonne National Lab. (ANL), Argonne, IL (United States); Weston, T. [Univ. of Oklahoma, Norman, OK (United States); Xie, Junqie [Argonne National Lab. (ANL), Argonne, IL (United States); Xu, L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Zaffaroni, E. [Univ. of Geneva (Switzerland); Zhang, M. [Univ. of Illinois, Urbana, IL (United States); Argelles, C. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Axani, S. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Conrad, J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Katori, T. [Queen Mary Univ. of London (United Kingdom); Noulai, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Mandalia, S. [Queen Mary Univ. of London (United Kingdom); Sandstrom, P. [Univ. of Wisconsin, Madison, WI (United States); Kryemadhi, A. [Messiah College, Mechanicsburg, PA (United States); Barner, L. [Messiah College, Mechanicsburg, PA (United States); Grove, A. [Messiah College, Mechanicsburg, PA (United States); Mohler, J. [Messiah College, Mechanicsburg, PA (United States); Roth, A. [Messiah College, Mechanicsburg, PA (United States); Beuzekom, M. van [Nikhef National Inst. for Subatomic Physics, Amsterdam (Netherlands); Dall' Occo, E. [Nikhef National Inst. for Subatomic Physics, Amsterdam (Netherlands); Schindler, H. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Paley, J. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Badgett, W. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Denisov, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Lukic, S. [Vinca Inst. of Nuclear Sciences, Belgrade (Serbia); Ujic, P. [Vinca Inst. of Nuclear Sciences, Belgrade (Serbia); Lebrun, P. L.G. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Fields, L. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Christian, D. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States); Zaki, R. [Radboud Univ., Nijmegen (Netherlands)

    2018-01-23

    This Technical Memorandum (TM) summarizes the Fermilab Test Beam operations for FY2017. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the individual experiments that ran at FTBF and are listed in Table 1. Each experiment section was prepared by the relevant authors, and was edited for inclusion in this summary.

  3. Fermilab Testbeam Facility Annual Report – FY 2015

    Energy Technology Data Exchange (ETDEWEB)

    Albrow, M. G. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States)

    2015-11-01

    This Technical Memorandum (TM) summarizes the Fermilab Test Beam operations for FY 2015. It is one of a series of annual publications intended to gather information in one place. In this case, the information concerns the individual experiments that ran at FTBF and are listed in Table TB-1. Each experiment section was prepared by the relevant authors, and was edited for inclusion in this summary.

  4. RA research nuclear reactor - Annual report 1985; Istrazivacki nuklearni reaktor RA - Izvestaj za 1985. godinu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-12-01

    Research reactor RA Annual report for year 1985 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1985 godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  5. RA Research reactor, Annual report 1986; Istrazivacki nuklearni reaktor RA - Izvestaj za 1986. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1986-12-01

    Research reactor RA Annual report for year 1985 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1986 godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA.

  6. Determination of Source Term for an Annual Stack Release of Gas Reactor G.A. Siwabessy

    International Nuclear Information System (INIS)

    Sudiyati; Syahrir; Unggul Hartoyo; Nugraha Luhur

    2008-01-01

    Releases of radionuclide from the reactor are noble gases, halogenides and particulates. The measurements were carried out directly on the air monitoring system of the stack. The results of these measurements are compared with the annual Source-Term data from the Safety Analyses report (SAR) of RSG-GAS. The measurement results are smaller than the data reported in SAR document. (author)

  7. Umatilla hatchery satellite facilities operation and maintenance. Annual report 1996

    International Nuclear Information System (INIS)

    Rowan, G.D.

    1997-06-01

    The Confederated Tribes of the Umatilla Indian Reservation (CTUIR) and Oregon Department of Fish and Wildlife (ODFW) are cooperating in a joint effort to enhance steelhead and re-establish salmon runs in the Umatilla River Basin. As an integral part of this program, Bonifer Pond, Minthorn Springs, Imeques C-mem-ini-kem and Thornhollow satellite facilities are operated for acclimation and release of juvenile summer steelhead (Oncorhynchus mykiss), fall and spring chinook salmon (O. tshawytscha) and coho salmon (O. kisutch). Minthorn is also used for holding and spawning adult summer steelhead and Three Mile Dam is used for holding and spawning adult fall chinook and coho salmon. Bonifer, Minthorn, Imeques and Thornhollow facilities are operated for acclimation and release of juvenile salmon and summer steelhead. The main goal of acclimation is to reduce stress from trucking prior to release and improve imprinting of juvenile salmonids in the Umatilla River Basin. Juveniles are transported to the acclimation facilities primarily from Umatilla and Bonneville Hatcheries. This report details activities associated with operation and maintenance of the Bonifer, Minthorn, Imeques, Thornhollow and Three Mile Dam facilities in 1996

  8. Fermilab Test Beam Facility Annual Report. FY 2014

    Energy Technology Data Exchange (ETDEWEB)

    Brandt, A. [Fermi National Accelerator Lab. (FNAL), Batavia, IL (United States). et al.

    2015-01-01

    Fermilab Test Beam Facility (FTBF) operations are summarized for FY 2014. It is one of a series of publications intended to gather information in one place. In this case, the information concerns the individual experiments that ran at FTBF. Each experiment section was prepared by the relevant authors, and was edited for inclusion in this summary.

  9. Annual report in compliance with the reactor sharing program, September 1, 1994--August 31, 1995

    International Nuclear Information System (INIS)

    Karam, R.A.

    1997-01-01

    This report contains information with regard to facilities utilization, descriptions (brief), personnel, organization, and programs of the Neely Nuclear Research Center (NNRC) at the Georgia Institute of Technology. The NNRC has two major facilities: the Georgia Tech Research Reactor and the Hot Cell Laboratory. This report of NNRC utilization is prepared in compliance with the contract requirements between the U.S. Department of Energy and the Georgia Institute of Technology. The NNRC is a participant in the University Reactor Sharing Program; as such, it makes available its 5 MW research reactor, its Co-60 irradiation facility and its activation analysis laboratory to large numbers of students and faculty from many universities and colleges

  10. Daily tritium intakes by people living near a heavy-water research reactor facility: dosimetric significance

    International Nuclear Information System (INIS)

    Trivedi, A.; Cornett, R.J.; Galeriu, D.; Workman, W.; Brown, R.M.

    1997-02-01

    We have estimated the relative daily intakes of tritiated water (HTO) and organically bound tritium (OBT), and have measured HTO-in-urine, in an adult population residing in the town of Deep River, Ontario, near a heavy-water research reactor facility at Chalk River. The daily intake of elevated levels of atmospheric tritium has been estimated from its concentration in environmental and biological samples, and various food items from a local tritium-monitoring program. Where the available data were inadequate, we used estimates generated by an environmental tritium-transfer model. From these data and estimates, we calculated a total daily tritium intake of about 55 Bq. Of this amount, 2.5 Bq is obtained from OBT-in-diet. Inhalation of HTO-in-air (15 Bq d -1 ) and HTO-in-drinking water (15 Bq d -1 ) accounts for more than half of the HTO intake. Skin absorption of HTO from air and bathing or swimming (for 30 min d -1 ) accounts for another 9 Bq d -1 and 0.1 Bq d -1 , respectively. The remaining intake of HTO is from food as tissue-free water tritium. The International Commission on Radiological Protection's recommended two-compartment metabolic model for tritium predicts an equilibrium body burden of about 900 Bq from HTO (818 Bq) and OBT (83 Bq) in the body, which corresponds to an annual tritium dose of 0.41 μSv. The model-predicted urinary excretion of HTO (∼18 Bq L -1 ) agrees well with measured HTO-in-urine (range, 10-32 Bq L -1 ). The OBT dose contribution to the total tritium dose is about 16%. We conclude that for the people living near the Chalk River research reactor facility, the bulk of the tritium dose is due to HTO intake. (author)

  11. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    1975-12-01

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

  12. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection.

  13. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  14. Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

    International Nuclear Information System (INIS)

    Brooks, B.G.

    1985-03-01

    This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

  15. Atomics International environmental monitoring and facility effluent annual report, 1976

    International Nuclear Information System (INIS)

    Moore, J.D.

    1977-01-01

    Environmental and facility effluent radioactivity monitoring at Atomics International (AI) is performend by the Radiation and Nuclear Safety Unit of the Health, Safety, and Radiation Services Department. Soil, vegetation, and surface water are routinely sampled to a distance of 10 miles from AI sites. Continuous ambient air sampling and thermoluminescent dosimetry are performed on site for monitoring airborne radioactivity and site ambient radiation levels. Radioactivity in effluents discharged to the atmosphere from AI facilities is continuously sampled and monitored to ensure that levels released to unrestricted areas are within appropriate limits, and to identify processes which may require additional engineering safeguards to minimize radioactivity levels in such effluents. In addition, selected nonradioactive constituents in surface water discharged to unrestricted areas are determined. This report summarizes and discusses monitoring results for 1976. The results of a special soil plutonium survey performed during the year are also summarized

  16. Status of safety at Areva group facilities. 2007 annual report

    International Nuclear Information System (INIS)

    2007-01-01

    This report describes the status of nuclear safety and radiation protection in the facilities of the AREVA group and gives information on radiation protection in the service operations, as observed through the inspection programs and analyses carried out by the General Inspectorate in 2007. Having been submitted to the group's Supervisory Board, this report is sent to the bodies representing the personnel. Content: 1 - A look back at 2007 by the AREVA General Inspector: Visible progress in 2007, Implementation of the Nuclear Safety Charter, Notable events; 2 - Status of nuclear safety and radiation protection in the nuclear facilities and service operations: Personnel radiation protection, Event tracking, Service operations, Criticality control, Radioactive waste and effluent management; 3 - Performance improvement actions; 4 - Description of the General Inspectorate; 5 - Glossary

  17. International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

    International Nuclear Information System (INIS)

    1973-01-01

    The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

  18. Feasibility study to develop BNCT facility at the Indonesian research reactor

    International Nuclear Information System (INIS)

    Hastowo, H.

    2001-01-01

    A survey on the Indonesian research reactors and its supporting facilities has been done in order to check the possibility to install BNCT facility. Oncologists from several hospitals have been informing about the BNCT treatment for tumours and they give a positive response to support utilisation of the BNCT facility. Several aspects required to support the BNCT treatment have also been identified and related activities on that matter soon will be initiated. The interim result in our survey indicated that utilisation of the 30 MW Multipurpose reactor would not be possible from the technical point of view. Further study will be concentrated on the TRIGA reactor and an epithermal neutron beam facility at the thermal column of this reactor will be designed for further work. (author)

  19. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: - To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; - To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; - To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; - To discuss the results of studies and on-going R&D activities that address cost reduction and the future economic competitiveness of fast reactors; and - To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  20. Technical Meeting on Fast Reactors and Related Fuel Cycle Facilities with Improved Economic Characteristics. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objectives of the meeting were: • To identify the main issues and technical features that affect capital and energy production costs of fast reactors and related fuel cycle facilities; • To present fast reactor concepts and designs with enhanced economic characteristics, as well as innovative technical solutions (components, subsystems, etc.) that have the potential to reduce the capital costs of fast reactors and related fuel cycle facilities; • To present energy models and advanced tools for the cost assessment of innovative fast reactors and associated nuclear fuel cycles; • To discuss the results of studies and ongoing R&D activities that address cost reduction and the future economic competitiveness of fast reactors; • To identify research and technology development needs in the field, also in view of new IAEA initiatives to help and support Member States in improving the economic competitiveness of fast reactors and associated nuclear fuel cycles

  1. Alternative water injection device to reactor equipment facility

    International Nuclear Information System (INIS)

    Yamashita, Masahiro.

    1995-01-01

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

  2. A review of experiments and results from the transient reactor test (TREAT) facility

    International Nuclear Information System (INIS)

    Deitrich, L. W.

    1998-01-01

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

  3. Ohmically heated toroidal experiment (OHTE) mobile ignition test reactor facility concept study

    International Nuclear Information System (INIS)

    Masson, L.S.; Watts, K.D.; Piscitella, R.R.; Sekot, J.P.; Drexler, R.L.

    1983-02-01

    This report presents the results of a study to evaluate the use of an existing nuclear test complex at the Idaho National Engineering Laboratory (INEL) for the assembly, testing, and remote maintenance of the ohmically heated toroidal experiment (OHTE) compact reactor. The portable reactor concept is described and its application to OHTE testing and maintenance requirements is developed. Pertinent INEL facilities are described and several test system configurations that apply to these facilities are developed and evaluated

  4. Mathematical modelling and quality indices optimization of automatic control systems of reactor facility

    International Nuclear Information System (INIS)

    Severin, V.P.

    2007-01-01

    The mathematical modeling of automatic control systems of reactor facility WWER-1000 with various regulator types is considered. The linear and nonlinear models of neutron power control systems of nuclear reactor WWER-1000 with various group numbers of delayed neutrons are designed. The results of optimization of direct quality indexes of neutron power control systems of nuclear reactor WWER-1000 are designed. The identification and optimization of level control systems with various regulator types of steam generator are executed

  5. Berkeley Nuclear Laboratories Reactor Physics Mk. III Experimental Programme. Description of facility and programme for 1971

    Energy Technology Data Exchange (ETDEWEB)

    Nunn, R M; Waterson, R H; Young, J D

    1971-01-15

    Reactor physics experiments have been carried out at Berkeley Nuclear Laboratories during the past few years in support of the Civil Advanced Gas-Cooled Reactors (Mk. II) the Generating Board is building. These experiments are part of an overall programme whose objective is to assess the accuracy of the calculational methods used in the design and operation of these reactors. This report provides a description of the facility for the Mk. III experimental programme and the planned programme for 1971.

  6. Annual report 1987-1988

    International Nuclear Information System (INIS)

    1988-01-01

    The paper presents: the annual report 1987/8 of the University Research Reactor, Risley, United Kingdom. The report contains a description of the current research programme using reactor produced isotopes and other reactor facilities. A summary of the work carried out by the Activation Analysis Service during 1987/8 is also given, along with the Reactor Teaching Programme, and the reactor operation and site safety arrangements. (U.K.)

  7. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  8. Operating manual for the High Flux Isotope Reactor. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    None

    1965-06-01

    This report contains a comprehensive description of the High Flux Isotope Reactor facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procedures are presented in another report.

  9. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    International Nuclear Information System (INIS)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report

  10. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  11. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  12. Annual report to the Laser Facility Committee 1978

    International Nuclear Information System (INIS)

    1978-01-01

    The report is in sections, as follows: the development of the facility (glass laser physics and development, performance and reliability of the glass laser, computer control, target fabrication, target area, optical design, gas laser development); single beam interaction studies (optical and magnetic measurements, X-ray and VUV spectroscopy, optical emission studies, particle emission measurements, gas breakdown observations, related theoretical and computational studies); two beam compression studies (vacuum ultra violet and X-ray spectroscopy, optical spectroscopy, particle emission studies, optical and magnetic measurements, theory and computational modelling). (U.K.)

  13. Annual report of the research works with joint-use JAERI facilities for fiscal 1974

    International Nuclear Information System (INIS)

    1975-01-01

    Results of the research works by national universities with JAERI's (Japan Atomic Energy Research Institute) joint-use facilities for fiscal 1974 are described. Facilities are research reactors, Co-60, Linac, etc. Research results are presented in individual summaries, covering radiation damage and solid state physics, activation analysis and nuclear chemistry, irradiation effects, etc. Results of the joint works with JAERI are also presented similarly. (Mori, K.)

  14. Atmospheric Radiation Measurement Climate Research Facility (ACRF) Annual Report 2008

    Energy Technology Data Exchange (ETDEWEB)

    LR Roeder

    2008-12-01

    The Importance of Clouds and Radiation for Climate Change: The Earth’s surface temperature is determined by the balance between incoming solar radiation and thermal (or infrared) radiation emitted by the Earth back to space. Changes in atmospheric composition, including greenhouse gases, clouds, and aerosols, can alter this balance and produce significant climate change. Global climate models (GCMs) are the primary tool for quantifying future climate change; however, there remain significant uncertainties in the GCM treatment of clouds, aerosol, and their effects on the Earth’s energy balance. In 1989, the U.S. Department of Energy (DOE) Office of Science created the Atmospheric Radiation Measurement (ARM) Program to address scientific uncertainties related to global climate change, with a specific focus on the crucial role of clouds and their influence on the transfer of radiation in the atmosphere. To reduce these scientific uncertainties, the ARM Program uses a unique twopronged approach: • The ARM Climate Research Facility, a scientific user facility for obtaining long-term measurements of radiative fluxes, cloud and aerosol properties, and related atmospheric characteristics in diverse climate regimes; and • The ARM Science Program, focused on the analysis of ACRF and other data to address climate science issues associated with clouds, aerosols, and radiation, and to improve GCMs. This report provides an overview of each of these components and a sample of achievements for each in fiscal year (FY) 2008.

  15. Computer simulation of radiation processes in reactor facilities

    International Nuclear Information System (INIS)

    Gann, V.V.; Abdulaev, A.M.; Zhukov, A.I.; Marekhin, S.V.; Soldatov, S.A.

    2009-01-01

    The paper describes experience of the code system ALPHA-H/PHOENIX-H/ANC-H (APA) and the code MCNP usage for fuel assembly neutronic calculations and modeling of VVER-1000 reactor core. Using Monte Carlo code MCNP, calculations of neutron field and pin-by-pin energy deposition distributions are provided for different type of assemblies in reactor core. An MCNP model for unit 3 Zaporozhye NPP reactor core was designed. Calculations for pin-by-pin energy deposition in the reactor core were performed using the code system APA and the code MCNP. Comparison of these calculations shows rather high precision of APA calculation for energy deposition in the fuel rods and assemblies operated in the reactor core

  16. Field-reversed mirror pilot reactor. Annual report

    International Nuclear Information System (INIS)

    Devoto, R.S.; Erickson, J.L.; Fink, J.H.

    1980-09-01

    This report concludes a two-year effort to design a near-term small-scale fusion power plant which, through its construction and operation, would be a direct and important step toward the commercialization of fusion energy. The fusion reactor pilot plant was designed under the ground rules that it must produce net power, be compact, have minimum total cost, and use near-term (late 1980's) engineering technology. The neutral beam driven, field-reversed mirror (FRM) was selected as the fusion plasma confinement concept around which the pilot plant was designed. Although the physics data base for this design is not yet well in hand, it is being pursued within the magnetic field-reversal framework of the US Mirror Fusion Program. Depending on the plasma size, the pilot plant would gross up to 19.8 MW(e) and would produce up to 10.7 MW(e) net, with the recirculated power used principally for the neutral beam injectors and refrigeration for the superconducting magnets

  17. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    International Nuclear Information System (INIS)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters

  18. Education and research at the VR-1 Vrabec training reactor facility

    International Nuclear Information System (INIS)

    Matejka, K.

    1993-01-01

    The results of 12 years' efforts devoted to the construction of the VR-1 ''Vrabec'' training reactor at the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague and to establishing the training reactor department, as well as the contribution of the training reactor facility to the teaching and scientific activities of the Faculty are presented in a comprehensive manner. The thesis is divided into 2 parts: (i) preconditions, reactor construction and commissioning, and constituting the reactor department, and (ii) basic and comprehensive information concerning the current utilization of the reactor for the benefit of students from various university level institutions. The prospects of scientific activities of the department are also outlined. Attention is paid to selected nuclear safety aspects of the reactor during operation and teaching of students, as well as to its innovated digital control system whose implementation is planned. The results achieved are compared with the initial goals and with similar experience abroad. (P.A.)

  19. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion

    International Nuclear Information System (INIS)

    Marques, J.G.; Sousa, M.; Santos, J.P.; Fernandes, A.C.

    2011-01-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1 MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  20. Environmental assessment for the deactivation of the N Reactor facilities. Revision 1

    International Nuclear Information System (INIS)

    1994-11-01

    This environmental assessment (EA) provides information for the US Department of Energy (DOE) to decide whether the Proposed Action for the N Reactor facilities warrants a Finding of No Significant Impact or requires the preparation of an environmental impact statement (EIS). The EA describes current conditions at the N Reactor facilities, the need to take action at the facilities, the elements of the Proposed Action and alternatives, and the potential environmental impacts. The N Reactor facilities are currently in a surveillance and maintenance program, and will eventually be decontaminated and decommissioned (D and D). Operation and maintenance of the facilities resulted in conditions that could adversely impact human health or the environment if left as is until final D and D. The Proposed Action would deactivate the facilities to remove the conditions that present a potential threat to human health and the environment and to reduce surveillance and maintenance requirements. The action would include surveillance and maintenance after deactivation. Deactivation would take about three years and would involve about 80 facilities. Surveillance and maintenance would continue until final D and D, which is expected to be complete for all facilities except the N Reactor itself by the year 2018

  1. Radiological considerations in the design of Reprocessing Uranium Plant (RUP) of Fast Reactor Fuel Cycle Facility (FRFCF), Kalpakkam

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    A Fast Reactor Fuel Cycle Facility (FRFCF) being planned at Indira Gandhi Centre for Atomic Research, Kalpakkam is an integrated facility with head end and back end of fuel cycle plants co-located in a single place, to meet the refuelling needs of the prototype fast breeder reactor (PFBR). Reprocessed uranium oxide plant (RUP) is one such plant in FRFCF to built to meet annual requirements of UO 2 for fabrication of fuel sub-assemblies (FSAs) and radial blanket sub-assemblies (RSAs) for PFBR. RUP receives reprocessed uranium oxide powder (U 3 O 8 ) from fast reactor fuel reprocessing plant (FRP) of FRFCF. Unlike natural uranium oxide plant, RUP has to handle reprocessed uranium oxide which is likely to have residual fission products activity in addition to traces of plutonium. As the fuel used for PFBR is recycled within these plants, formation of higher actinides in the case of plutonium and formation of higher levels of 232 U in the uranium product would be a radiological problem to be reckoned with. The paper discussed the impact of handling of multi-recycled reprocessed uranium in RUP and the radiological considerations

  2. An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

    International Nuclear Information System (INIS)

    Forsbacka, Matt

    1990-01-01

    A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)

  3. RB research nuclear reactor, Annual report for 1984, I - III; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1984. godini, I - III

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Pesic, M; Vranic, S; Petronijevic, M; Zivkovic, B; Ilic, I [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1984-07-01

    The annual report for 1984 contains 3 parts. Part one includes the following: description of the reactor, exploitation possibilities of the reactor, reactor operation, accident and incidents analysis; reactor equipment and components; dosimetry and radiation protection; RB reactor staff and financial data. Part two of this report is devoted to maintenance and control of reactor components, electronic and electric equipment as well as auxiliary systems. Part three describes reactor exploitation; development of experimental methods; utilization of the reactor as a radiation source.

  4. A description of the Canadian irradiation-research facility proposed to replace the NRU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, A G; Lidstone, R F; Bishop, W E; Talbot, E F; McIlwain, H [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-31

    To replace the aging NRU reactor, AECL has developed the concept for a dual-purpose national Irradiation Research Facility (IRF) that tests fuel and materials for CANDU (CANada Deuterium Uranium) reactors and performs materials research using extracted neutron beams. The IRF includes a MAPLE reactor in a containment building, experimental facilities, and support facilities. At a nominal reactor power of 40 MW{sub t}, the IRF will generate powers up to 1 MW in natural-uranium CANDU bundles, fast-neutron fluxes up to 1.4 x 10{sup 18} n{center_dot}m{sup -2}{center_dot}s{sup -1} in Zr-alloy specimens, and thermal-neutron fluxes matching those available to the NRU beam tubes. (author). 9 refs., 5 tabs., 2 figs.

  5. 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

  6. Engineering evaluation/cost analysis for the 105-DR and 105-F Reactor facilities and ancillary facilities

    International Nuclear Information System (INIS)

    Coenenberg, E.T.

    1998-01-01

    This document presents the results of an engineering evaluation/cost analysis (EE/CA) that was conducted to evaluate alternatives to address final disposition of the 105-DR and 105-F Reactor Buildings (subsequently referred to as facilities), including the fuel storage basins (FSB) and below-grade portions of the reactors, excluding the reactor blocks. The reactor blocks will remain in a safe storage mode for up to 75 years as identified in the Record the Decision (ROD) (58 FR 48509) for the Environmental Impact Statement (EIS), Decommissioning of Eight Surplus Production Reactors at the Hanford Site, Richland, Washington (DOE 1992a). This EE/CA also addresses final disposition of four ancillary facilities: 116-D and 116-DR Exhaust Air Stacks, 117-DR Exhaust Filter Building, and 119-DR Exhaust Air Sample Building. The 105-DR and 105-F facilities are located in the 100-D and 100-F Areas of the Hanford Site. In November 1989, the 100 Area of the Hanford Site was placed on the U.S. Environmental Protection Agency's (EPA) National Priorities List (NPL) under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The 100 Area NPL includes the 100-D Area (which includes the 100-DR site) and the 100-F Area, which are in various stages of the remediation process. It has been determined by DOE that hazardous substances in the 105-DR, 105-F, and the four ancillary facilities may present a potential threat to human health or the environment, and that a non-time critical removal action at these facilities is warranted. To help determine the most appropriate action, DOE, in cooperation with the Washington State Department of Ecology (Ecology) and the EPA, has prepared this EE/CA. The scope of the evaluation includes the 105-DR and 105-F facilities and the four ancillary facilities. The 116-DR and 117-DR facilities are located within the boundaries of the 105-DR Large Sodium Fire Facility Treatment, Storage, and Disposal (TSD) unit, which is

  7. Situation of the radioactive waste management and the employee radiation exposure in commercial power generation reactor facilities in fiscal 1980

    International Nuclear Information System (INIS)

    1981-01-01

    (1) Situation of the radioactive waste management in commercial power generating reactor facilities: The owners of power generation reactor facilities are obligated not to exceed the target dose around the sites by law in the radioactive waste management. The release of radioactive gaseous and liquid wastes and the storage of radioactive solid wastes in respective reactor facilities in fiscal 1980 are presented in tables (for the former, the data since 1971 are also given). The release control values were satisfied in all the facilities. (2) Situation of employe radiation exposure in commercial power generating reactor facilities: The owners of power generation reactor facilities are obligated not to exceed the permissible exposure doses by law. The Employe exposure doses in respective reactor facilities in fiscal 1980 are given in tables. All exposure doses were below the permissible levels. (J.P.N.)

  8. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    International Nuclear Information System (INIS)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed

  9. A Design of Alarm System in a Research Reactor Facility

    International Nuclear Information System (INIS)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk

    2013-01-01

    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants

  10. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.; Ott, K.O.; Terry, W.K.

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

  11. Progress on ANSTO'S OPAL reactor project and its future importance as the centrepiece of ANSTO'S facilities

    International Nuclear Information System (INIS)

    Smith, I.O.

    2006-01-01

    Full text: After an intensive process of analysis, the Australian government approved the construction of a multi-purpose research reactor in 1997. Following the conduct of a comprehensive tender evaluation process in 1998-2000, INVAP was contracted to construct a 20 MW open pool research reactor and associated neutron beam facilities. The construction of the reactor is now almost complete, and we have commenced cold commissioning. ANSTO has applied for an operating licence, and we hope for a decision on that application in June, following the consideration by the regulator of the results of cold commissioning. The OPAL reactor will provide neutrons to a world-class neutron beam facility, in which a number of the instruments will have the best performance available in the world to date. We intend to establish the Bragg Institute as a regional centre of excellence on neutron beam science, with a significant number of international scientists using the facility to produce cutting edge science in the fields of biology, materials science, food science and other area. The reactor also has extensive irradiation facilities within the reflector vessel. These facilities will be used to produce medical isotopes - ANSTO supplies the bulk of the Australian market and also exports into this region - and for the transmutation doping of silicon ingots for semiconductor manufacture. There are also a number of pneumatically loaded radiation facilities allowing for short term irradiation of samples for such activities as neutron activation analysis

  12. The Argonne Leadership Computing Facility 2010 annual report.

    Energy Technology Data Exchange (ETDEWEB)

    Drugan, C. (LCF)

    2011-05-09

    Researchers found more ways than ever to conduct transformative science at the Argonne Leadership Computing Facility (ALCF) in 2010. Both familiar initiatives and innovative new programs at the ALCF are now serving a growing, global user community with a wide range of computing needs. The Department of Energy's (DOE) INCITE Program remained vital in providing scientists with major allocations of leadership-class computing resources at the ALCF. For calendar year 2011, 35 projects were awarded 732 million supercomputer processor-hours for computationally intensive, large-scale research projects with the potential to significantly advance key areas in science and engineering. Argonne also continued to provide Director's Discretionary allocations - 'start up' awards - for potential future INCITE projects. And DOE's new ASCR Leadership Computing (ALCC) Program allocated resources to 10 ALCF projects, with an emphasis on high-risk, high-payoff simulations directly related to the Department's energy mission, national emergencies, or for broadening the research community capable of using leadership computing resources. While delivering more science today, we've also been laying a solid foundation for high performance computing in the future. After a successful DOE Lehman review, a contract was signed to deliver Mira, the next-generation Blue Gene/Q system, to the ALCF in 2012. The ALCF is working with the 16 projects that were selected for the Early Science Program (ESP) to enable them to be productive as soon as Mira is operational. Preproduction access to Mira will enable ESP projects to adapt their codes to its architecture and collaborate with ALCF staff in shaking down the new system. We expect the 10-petaflops system to stoke economic growth and improve U.S. competitiveness in key areas such as advancing clean energy and addressing global climate change. Ultimately, we envision Mira as a stepping-stone to exascale-class computers

  13. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    International Nuclear Information System (INIS)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han

    2013-01-01

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility

  14. Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.

  15. The upgrading of the cyclic neutron activation analysis facility at the Dalat research reactor

    International Nuclear Information System (INIS)

    Van Doanh Ho; Manh Dung Ho; Quang Thien Tran; Dong Vu Cao; Thanh Viet Ha

    2018-01-01

    The cyclic neutron activation analysis (CNAA) facility based on a pneumatic transfer system for short irradiation and rapid counting has recently been upgraded at the Dalat research reactor. The original facility was only designed for single irradiation. Therefore, this work has aimed to upgrade both hardware and software for the cyclic irradiation. In this paper, the upgrading of the facility for CNAA was described. Irradiation time of the facility were calibrated, thereby reducing irradiation time to seconds with precision. The accuracy and sensitivity of CNAA based-on the upgraded facility were assessed by determination of some short-lived nuclides. (author)

  16. Public's right to information: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Stokely, E.

    1981-02-01

    The events at TMI prompted the Under Secretary of the Department of Energy (DOE) to establish the Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee. This Committee was assigned the task of assessing the adequacy of nuclear facility personnel qualification and training at DOE-owned reactors in light of the Three Mile Island accident. The Committee was also asked to review recommendations and identify possible implications for DOE's nuclear facilities

  17. NRX and NRU reactor research facilities and irradiation and examination charges

    International Nuclear Information System (INIS)

    1960-08-01

    This report details the irradiation and examination charges on the NRX and NRU reactors at the Chalk River Nuclear Labs. It describes the NRX and NRU research facilities available to external users. It describes the various experimental holes and loops available for research. It also outlines the method used to calculate the facilities charges and the procedure for applying to use the facilities as well as the billing procedures.

  18. Technical committee meeting on Liquid Metal Fast Reactor (LMFR) developments. 33rd annual meeting of the International Working Group on Fast Reactors (IWG-FR). Working material

    International Nuclear Information System (INIS)

    2000-01-01

    Over the past 33 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1999/2000, as reported at the 33. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

  19. International Working Group on Fast Reactors Thirteenth Annual Meeting. Summary Report. Part II

    International Nuclear Information System (INIS)

    1980-10-01

    The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

  20. International Working Group on Fast Reactors Thirteenth Annual Meeting. Summary Report. Part I

    International Nuclear Information System (INIS)

    1980-09-01

    The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

  1. International Working Group on Past Reactors Thirteenth Annual Meeting. Summary Report. Part III

    International Nuclear Information System (INIS)

    1981-04-01

    The Thirteenth Annual Meeting of the IAEA International Working Group on Fast Reactors was held at the IAEA Headquarters, Vienna, Austria from 9 to 11 April 1980. The Summary Report (Part I) contains the Minutes of the Meeting. The Summary Report (Part II) contains the papers which review the national programme in the field of LMFBRs and other presentations at the Meeting. The Summary Report (Part III) contains the discussions on the review of the national programmes

  2. 2016 Annual Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Ponds

    International Nuclear Information System (INIS)

    Lewis, Michael George

    2017-01-01

    This report describes conditions and information, as required by the state of Idaho, Department of Environmental Quality Reuse Permit I-161-02, for the Advanced Test Reactor Complex Cold Waste Ponds located at Idaho National Laboratory from November 1, 2015-October 31, 2016. The effective date of Reuse Permit I-161-02 is November 20, 2014 with an expiration date of November 19, 2019. This report contains the following information: · Facility and system description · Permit required effluent monitoring data and loading rates · Permit required groundwater monitoring data · Status of compliance activities · Issues · Discussion of the facility's environmental impacts. During the 2016 permit year, 180.99 million gallons of wastewater were discharged to the Cold Waste Ponds. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest in well USGS-065, which is the closest downgradient well to the Cold Waste Ponds. Sulfate and total dissolved solids concentrations decrease rapidly as the distance downgradient from the Cold Waste Ponds increases. Although concentrations of sulfate and total dissolved solids are significantly higher in well USGS-065 than in the other monitoring wells, both parameters remained below the Ground Water Quality Rule Secondary Constituent Standards in well USGS-065. The facility was in compliance with the Reuse Permit during the 2016 permit year.

  3. Argonne Leadership Computing Facility 2011 annual report : Shaping future supercomputing.

    Energy Technology Data Exchange (ETDEWEB)

    Papka, M.; Messina, P.; Coffey, R.; Drugan, C. (LCF)

    2012-08-16

    The ALCF's Early Science Program aims to prepare key applications for the architecture and scale of Mira and to solidify libraries and infrastructure that will pave the way for other future production applications. Two billion core-hours have been allocated to 16 Early Science projects on Mira. The projects, in addition to promising delivery of exciting new science, are all based on state-of-the-art, petascale, parallel applications. The project teams, in collaboration with ALCF staff and IBM, have undertaken intensive efforts to adapt their software to take advantage of Mira's Blue Gene/Q architecture, which, in a number of ways, is a precursor to future high-performance-computing architecture. The Argonne Leadership Computing Facility (ALCF) enables transformative science that solves some of the most difficult challenges in biology, chemistry, energy, climate, materials, physics, and other scientific realms. Users partnering with ALCF staff have reached research milestones previously unattainable, due to the ALCF's world-class supercomputing resources and expertise in computation science. In 2011, the ALCF's commitment to providing outstanding science and leadership-class resources was honored with several prestigious awards. Research on multiscale brain blood flow simulations was named a Gordon Bell Prize finalist. Intrepid, the ALCF's BG/P system, ranked No. 1 on the Graph 500 list for the second consecutive year. The next-generation BG/Q prototype again topped the Green500 list. Skilled experts at the ALCF enable researchers to conduct breakthrough science on the Blue Gene system in key ways. The Catalyst Team matches project PIs with experienced computational scientists to maximize and accelerate research in their specific scientific domains. The Performance Engineering Team facilitates the effective use of applications on the Blue Gene system by assessing and improving the algorithms used by applications and the techniques used to

  4. Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

    International Nuclear Information System (INIS)

    2002-02-01

    The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

  5. Annual report of department of research reactor, 2000. April 1, 2000 - March 31, 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-02-01

    The Department of Research Reactor is responsible for the operation, Maintenance, utilization of the JRR-3 and the JRR-4 and for the related R and D. Besides RI production including its R and D are carried out. This report describes the activities of the department in fiscal year of 2000 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. RI Production and its R and D works were conducted as well. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for research reactors. Although the term 'JRR-3M' was used to denote the JRR-3M modified 1990 until the 2000 annual report of the Department of Research Reactor, the term 'JRR-3' will be used from this annual report because the JRR-3 has been operated for about 10 years since the modification and is now under further modification and upgrading study. (author)

  6. Implementation of safeguards and security for fissile materials disposition reactor alternative facilities

    International Nuclear Information System (INIS)

    Jaeger, C.D.; Duggan, R.A.; Tolk, K.M.

    1995-01-01

    A number of different disposition alternatives are being considered and include facilities which provide for long-ten-n and interim storage, convert and stabilize fissile materials for other disposition alternatives, immobilize fissile material in glass and/or ceramic material, fabricate fissile material into mixed oxide (MOX) fuel for reactors, use reactor based technologies to convert material into spent fuel, and dispose of fissile material using a number of geologic alternatives. Particular attention will be given to the reactor alternatives which include existing, partially completed, advanced or evolutionary LWRs and CANDU reactors. The various reactor alternatives are all very similar and include processing which converts Pu to a usable form for fuel fabrication, a MOX fuel fab facility located in either the US or in Europe, US LWRs or the CANDU reactors and ultimate disposal of spent fuel in a geologic repository. This paper focuses on how the objectives of reducing security risks and strengthening arms reduction and nonproliferation will be accomplished and the possible impacts of meeting these objectives on facility operations and design. Some of the areas in this paper include: (1) domestic and international safeguards requirements, (2) non-proliferation criteria and measures, (3) the threat, and (4) potential proliferation risks, the impacts on the facilities, and safeguards and security issues unique to the presence of Category 1 or strategic special nuclear material

  7. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    International Nuclear Information System (INIS)

    Bissani, M; O'Kelly, D S

    2006-01-01

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to provide color-enhanced gemstones but is

  8. Operation and maintenance of the RA reactor, RA Research reactor. Annual report 1976

    International Nuclear Information System (INIS)

    Martinc, R.

    1976-12-01

    During 1976 the Ra reactor was operating for about 30% shorter period than usual. The reason were extraordinary repair activities within regular and investment maintenance as well as repair of failures caused by neglected maintenance during previous 6 years. Delay was caused by unavailability of fuel (2% enriched fuel elements are spent) and the new 80% enriched fuel demanded experimental and theoretical analyses before being introduced into the core. Safety analyses concerned with using 80% enriched fuel both experimental and theoretical were successfully fulfilled. The December 1976 successful experimental campaign can be marked as end of the 17 years period of using 2% enriched fuel and start of the new period of using highly enriched fuel. This is significant not only for the reactor itself but for the users, because it would result in increase of neutron flux by 50% with the increase of costs by only 4%. Demand was submitted for obtaining the final license for transition operating regime with highly enriched fuel which would save at least 2 200 000 dinars. This will enable reactor operation in 1977 and later on, without interruption by 'critical' and other experiments related to new highly enriched fuel. A high number of repair and other urgent activities were fulfilled in order to enable safe operation. Some of these activities were done never before and some were neglected during past 6 years. The most important tasks were: purchase of Al tubes made of special alloy, fabrication and mounting of the fuel channel; overall investigation of reactor vessel leakage; repair of the heavy water pump; exchange of two vertical channels. basic equipment for construction of emergency cooling system was purchased. Hot cells are equipped for independent utilisation [sr

  9. Minthorn Springs Creek summer juvenile release and adult collection facility : annual report 1990.; ANNUAL

    International Nuclear Information System (INIS)

    Lofy, Peter T.; Rowan, Gerald D.

    1991-01-01

    The Confederated Tribes of the Umatilla Indian Reservation (CTUIR) and Oregon Department of Fish and Wildlife (ODFW) are cooperating in a joint effort to increase steelhead and re-establish salmon runs in the Umatilla River Basin. As part of this program, Bonifer and Minthorn Acclimation Facilities are operated for holding and spawning adult steelhead and acclimation and release of juvenile salmon and steelhead. Regularly-scheduled maintenance was completed in 1990. Equipment and pumps received maintenance and repair. Two of the Minthorn and all of the Bonifer pond outlet screens were replaced with vertical bars to alleviate clogging problems. A horizontal bar screen was installed in the water control structure at the largest spring at Bonifer to prevent fish from migrating upstream during acclimation. A pipe was installed under the railroad tracks at Bonifer to make unloading of fish from transport trucks easier and safer. The Minthorn access road was repaired to provide better access for delivery of fish to the facility and for general operations and maintenance

  10. Fast reactor test facilities in the US safety program

    International Nuclear Information System (INIS)

    Avery, R.; Dickerman, C.E.; Lennox, D.H.; Rose, D.

    1979-01-01

    The needs for safety information derivable from in-pile programs are reviewed, and the correlation made with existing and planned capability. In view of the current status of the U.S. breeder program, emphasis is given in the review to the impact of different fast breeder options on the required program and facilities. It is concluded that facility needs are somewhat independent of specific fast breeder concept, even though the relative emphasis on the various safety issues will differ. 8 refs

  11. Annual Status Report (FY2016) Performance Assessment for the Environmental Restoration Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Casbon, M. A. [CH2M HILL Plateau Remediation Company, Richland, WA (United States); Nichols, W. E. [CH2M HILL Plateau Remediation Company, Richland, WA (United States)

    2017-03-15

    DOE O 435.1, Radioactive Waste Management, and DOE M 435.1-1, Radioactive Waste Management Manual, require that a determination of continued adequacy of the performance assessment (PA), composite analysis (CA), and disposal authorization statement (DAS) be made on an annual basis, and it must consider the results of data collection and analysis from research, field studies, and monitoring. Annual summaries of low-level waste (LLW) disposal operations must be prepared with respect to the conclusions and recommendations of the PA and CA, and a determination of the need to revise the PA or CA must be made. The annual summary requirement provides a structured approach for demonstrating the continued adequacy of the PA and CA in demonstrating a reasonable expectation that the performance objectives will be met. This annual summary addresses only the status of the Environmental Restoration Disposal Facility (ERDF) PA (CP-60089, Performance Assessment for the Environmental Restoration Disposal Facility, Hanford Site, Washington, formerly WCH-520 Rev. 1)1. The CA for ERDF is supported by DOE/RL-2016-62, Annual Status Report (FY 2016): Composite Analysis of Low Level Waste Disposal in the Central Plateau at the Hanford Site. The ERDF PA portion of the CA document is found in Section 3.1.4, and the ERDF operations portion is found in Section 3.3.3.2 of that document.

  12. Experimental measurements and theoretical simulations for neutron flux in self-serve facility of Dhruva reactor

    International Nuclear Information System (INIS)

    Rana, Y.S.; Mishra, Abhishek; Singh, Tej

    2016-06-01

    Dhruva is a 100 MW th tank type research reactor with natural metallic uranium as fuel and heavy water as coolant, moderator and reflector. The reactor is utilized for production of a large variety of radioisotopes for fulfilling growing demands of various applications in industrial, agricultural and medicinal sectors, and neutron beam research in condensed matter physics. The core consists of two on-power tray rods for radioisotope production and fifteen experimental beam holes for neutron beam research. Recently, a self-serve facility has also been commissioned in one of the through tubes in the reactor for carrying out short term irradiations. To get accurate information about neutron flux spectrum, measurements have been carried out in self-serve facility of Dhruva reactor. The present report describes measurement method, analysis technique and results. Theoretical estimations for neutron flux were also carried out and a comparison between theoretical and experimental results is made. (author)

  13. Annual monitoring and surveillance report for Piqua Nuclear Power Facility, Piqua, Ohio

    International Nuclear Information System (INIS)

    Mosho, G.D.

    1991-12-01

    This report discusses the decommissioned Piqua Nuclear Power Facility which is located in Piqua, Ohio near the Greater Miami River. The Facility was built by the US Atomic Energy Commission (now U. S. Department of Energy) and was operated from 1963 to 1966. The reactor was retired prior to 1970 and the facility was leased to the city of Piqua for use as offices and equipment storage. In December 1991, a radiological survey was done of the facility to document its radiological condition. The data show that all radiological parameters measured were essentially the same as that found in the natural environment. The only exception was that low levels of radioactive contamination were detected in one drain on the 56.5 ft elevation, but the radiation exposure rate in that area was also typical of natural background

  14. Operation and maintenance of the RB reactor, Annual report for 1977; Pogon i odrzavanje reaktora RB, Izvestaj o radu u 1977. godini

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O; Vranic, S [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1977-07-01

    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.

  15. RB research nuclear reactor, Annual report for 1981; Istrazivacki nuklearni reaktor RB, Izvestaj o radu u 1981. godini

    Energy Technology Data Exchange (ETDEWEB)

    Markovic, H; Sotic, O; Pesic, M; Vranic, S; Zivkovic, B; Bogdanovic, M; Petronijevic, M [Boris Kidric Institute of Nuclear Sciences Vinca, Beograd (Yugoslavia)

    1981-07-01

    The annual report for 1981 includes the following: utilization of the RB reactor; accident and incidents analysis; description of the reactor equipment status; dosimetry and radiation protection; RB reactor staff; financial data. Seven Annexes to this report are concerned with: maintenance of the reactor components and equipment, including nuclear fuel, heavy water, reactor vessel, heavy water coolant circuit, experimental platforms, absorption rods; maintenance of the electric power supply system, neutron source equipment, crane; control and maintenance of ventilation and heating systems, gas and comprised gas systems, fire protection system; plan for renewal of the reactor components; contents of the RB reactor safety report; reactor staff; review of measured radiation doses; experimental methods; training of the staff; and financial report.

  16. Status of national programmes on fast breeder reactors. Eighteenth annual meeting, Vienna, Austria, 16-19 April 1985

    International Nuclear Information System (INIS)

    1986-02-01

    The Eighteenth Annual Meeting on the Status of National Programmes in Member States of the IAEA on Fast Breeder Reactors had been held in April 1985. The representatives of the Member States and international organizations reported status and activities in the field of fast breeder reactors development and operation. A separate abstract was prepared for each of the 12 presentations of the meeting

  17. Emergency planning and response: An independent safety assessment of Department of Energy nuclear reactor facilities

    International Nuclear Information System (INIS)

    Knuth, D.; Boyd, R.

    1981-02-01

    The Department of Energy (DOE) has formed a Nuclear Facilities Personnel Qualification and Training (NFPQT) Committee to assess the implications of the recommendations contained in the President's Commission Report on the Three Mile Island (TMI) Accident (the Kemeny Commission report) that are applicable to DOE's nuclear reactor operations. Thirteen DOE nuclear reactors have been reviewed. The assessments of the 13 facilities are based on information provided by the individual operator organizations and/or cognizant DOE Field Offices. Additional clarifying information was supplied in some, but not all, instances. This report indicates how these 13 reactor facilities measure up in light of the Kemeny and other TMI-related studies and recommendations, particularly those that have resulted in upgraded Nuclear Regulatory Commission (NRC) requirements in the area of emergency planning and response

  18. Cold neutron PGAA facility developments at university research reactors in the USA

    International Nuclear Information System (INIS)

    Uenlue, K.; Rios-Martinez, C.

    2005-01-01

    The PGAA applications can be enhanced by using subthermal neutrons, cold neutrons at university research reactors. Only two cold neutron beam facilities were developed at the U.S. university research reactors, namely at Cornell University and the University of Texas at Austin. Both facilities used mesitylene moderator. The mesitylene moderator in the Cornell Cold Neutron Beam Facility (CNBF) was cooled by a helium cryorefrigerator via copper cold fingers to maintain the moderator below 30 K at full power reactor operation. Texas Cold Neutron Source (TCNS) also uses mesitylene moderator that is cooled by a cryorefrigerator via a neon thermosiphon. The operation of the TCNS is based on a helium cryorefrigerator, which liquefies neon gas in a 3-m long thermosiphon. The thermosiphon cools and maintains mesitylene moderator at about 30 K in a chamber. Neutrons streaming through the mesitylene chamber are moderated and thus reduce their energy to produce a cold neutron distribution. (author)

  19. Self-sustainability of a research reactor facility with neutron activation analysis

    International Nuclear Information System (INIS)

    Chilian, C.; Kennedy, G.

    2010-01-01

    Long-term self-sustainability of a small reactor facility is possible because there is a large demand for non-destructive chemical analysis of bulk materials that can only be achieved with neutron activation analysis (NAA). The Ecole Polytechnique Montreal SLOWPOKE Reactor Facility has achieved self-sustainability for over twenty years, benefiting from the extreme reliability, ease of use and stable neutron flux of the SLOWPOKE reactor. The industrial clientele developed slowly over the years, mainly because of research users of the facility. A reliable NAA service with flexibility, high accuracy and fast turn-around time was achieved by developing an efficient NAA system, using a combination of the relative and k0 standardisation methods. The techniques were optimized to meet the specific needs of the client, such as low detection limit or high accuracy at high concentration. New marketing strategies are presented, which aim at a more rapid expansion. (author)

  20. Accidents and failures in reactor facilities for test and research and reactor facilities in the stage of research and development in fiscal year 1987

    International Nuclear Information System (INIS)

    1988-01-01

    The number of accidents and failures reported in fiscal year 1987 in conformity with the law on the regulation of nuclear reactors and others was three. One case occurred during operation, and two cases occurred in shutdown state. One case was caused by improper construction management, and two cases were due to improper maintenance management. The effect of radioactivity to the surrounding environment of reactor facilities due to these accidents and failures did not arise. These occurred in the NSRR of Japan Atomic Energy Research Institute (Tokai), the experimental FBR Joyo and the ATR Fugen Power Station of Power Reactor and Nuclear Fuel Development Corp. In addition to these, the light troubles reported on the basis of the notice from the director of Science and Technology Agency dated September 1, 1981, were three cases. (K.I.)

  1. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    International Nuclear Information System (INIS)

    Harris, D.R.

    1995-01-01

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

  2. Experimental facility of innovative types as the laboratory analog of research reactor experimental device

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Zabud'ko, A.N.; Kremenetskij, A.K.; Nikolaev, A.N.; Trykov, L.A.

    1991-01-01

    The paper analyses capability of creating laboratory analogs of complex experimental facilities at research reactors utilizing power radionuclide neutron sources fabricated in industrial conditions. Some experimental and calculational investigations of neutron-physical characteristics are presented, which have been attained at the RIZ research reactor laboratory analog. Experimental results are supplemented by calculational investigations, fulfilled by means of the BRAND three-dimensional computational complex and the ROZ-6 one-dimensional program. 4 refs.; 3 figs

  3. The regulation and licensing of research reactors and associated facilities in the United Kingdom

    International Nuclear Information System (INIS)

    Weightman, M.W.; Willby, C.R.

    1990-01-01

    In the United Kingdom, the Nuclear Installations Inspectorate (NII) licenses nuclear facilities, including research reactors, on behalf of the Health and Safety Executive (HSE). The legislation, the regulatory organizations and the methods of operation that have been developed over the last 30 years result in a largely non-prescriptive form of control that is well suited to research reactors. The most important part of the regulatory system is the license and the attachment of conditions which it permits. These conditions require the licensee to prepare arrangements to control the safety of the facility. In doing so the licensee is encouraged to develop a 'safety culture' within its organization. This is particularly important for research reactors which may have limited staff resources and where the ability, and at times the need, to have access to the core is much greater than for nuclear power plants. Present day issues such as the ageing of nuclear facilities, public access to the rationale behind regulatory decisions, and the emergence of more stringent safety requirements, which include a need for quantified safety criteria, have been addressed by the NII. This paper explores the relevance of such issues to the regulation of research reactors. In particular, it discusses some of the factors associated with research reactors that should be considered in developing criteria for the tolerability of risk from these nuclear facilities. From a consideration of these factors, it is the authors' view that the range of tolerable risk to the public from the operation of new research reactors may be expected to be more stringent than similar criteria for new nuclear power plants, whereas the criteria for tolerable risk for research reactor workers are expected to be about the same as those for power reactor workers

  4. Procedures for conducting probabilistic safety assessment for non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    2002-01-01

    A well performed and adequately documented safety assessment of a nuclear facility will serve as a basis to determine whether the facility complies with the safety objectives, principles and criteria as stipulated by the national regulatory body of the country where the facility is in operation. International experience shows that the practices and methodologies used to perform safety assessments and periodic safety re-assessment for non-reactor nuclear facilities differ significantly from county to country. Most developing countries do not have methods and guidance for safety assessment that are prescribed by the regulatory body. Typically the safety evaluation for the facility is based on a case by case assessment. Whilst conservative deterministic analyses are predominantly used as a licensing basis in many countries, recently probabilistic safety assessment (PSA) techniques have been applied as a useful complementary tool to support safety decision making. The main benefit of PSA is to provide insights into the safety aspects of facility design and operation. PSA points up the potential environmental impacts of postulated accidents, including the dominant risk contributors, and enables safety analysts to compare options for reducing risk. In order to advise on how to apply PSA methodology for the safety assessment of non-reactor nuclear facilities, the IAEA organized several consultants meetings, which led to the preparation of this TECDOC. This document is intended as guidance for the conduct of PSA in non-nuclear facilities. The main emphasis here is on the general procedural steps of a PSA that is specific for a non-reactor nuclear facility, rather than the details of the specific methods. The report is directed at technical staff managing or performing such probabilistic assessments and to promote a standardized framework, terminology and form of documentation for these PSAs. It is understood that the level of detail implied in the tasks presented in this

  5. RA Research reactor, Annual report 1988; Istrazivacki nuklearni reaktor RA - Izvestaj za 1988. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1988-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1989. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjering RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  6. Research nuclear reactor RA - Annual Report 1989; Istrazivacki nuklearni reaktor RA - Izvestaj za 1989. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1989-12-15

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities. [Serbo-Croat] Godisnji izvestaj po projektu 'Istrazivacki nuklearni reaktor RA' za 1989. godinu, koji finansira republicka zajednica za nauku SR Srbije po ugovoru br. 3705/1 sastoji se iz dva dela. Prvi deo obuhvata pogon i odrzavanje nuklearnog reaktora RA, sto predstavlja obavezu Odeljenja za reaktorski inzenjering u sastavu OOUR Instituta za multidisciplinarna istrazivanja i inzenjering RA. Drugi deo obuhvata poslove zastite od zracenja na reaktoru RA, sto predstavlja obavezu OOUR Instituta za zastitu od zracenja 'Zastita'. Naucno vece Instituta za multidisciplinarna istrazivanja i inzenjering RA ocenilo je da sadrzina ovog izvestaja odgovara izvrsenim poslovima na reaktoru RA u 1989. godini. Pozitivno se ocenjuje obim i kvalitet izvrsenih radova kako u pogledu odrzavanja i rekonstrukcije reaktora, tako i u pogledu poslova zastite od zracenja izvrsenih kod njega.

  7. Annual report of Department of Research Reactor, 1997. April 1, 1997 - March 31, 1998

    International Nuclear Information System (INIS)

    1999-01-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of the department in fiscal year of 1997 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, the utilization of irradiation and neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and the department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  8. Annual report of department of research reactor, 1995 (April 1, 1995 - March 31, 1996)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1995 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  9. Annual report of Department of Research Reactor, 1996. April 1, 1996 - March 31, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The Department of Research Reactor is responsible for the operation, maintenance, utilization and related R and D works of the research reactors including JRR-2, JRR-3M (new JRR-3) and JRR-4. This report describes the activities of our department in fiscal year of 1996 and it also includes some of the technical topics on the works mentioned above. As for the research reactors, we carried out the operation, maintenance, irradiation utilization, neutron beam experiments, technical management including fuels and water chemistry, radiation monitoring as related R and D works. The international cooperations between the developing countries and our department were also made concerning the operation, utilization and safety analysis for nuclear facilities. (author)

  10. Annual report on JEN-1 and JEN-2 Reactors; Informe periodico de Reactores JEN-1 y JEN-2 correpondiente al ano 1972

    Energy Technology Data Exchange (ETDEWEB)

    Montes Ponce de Leon, J.

    1974-07-01

    In the annual report on the JEN-1 and JEN-2 reactors the main fractures of the reactor operations and maintenance are described. The reactor has been in operation for 2188 hours, what means 74% of the total working time. Maintenance and periodical tests have occupied the rest of the time. Maintenance operations are shown according to three main subjects, the main failures so as the reactor scrams are also described. Different date relating with radiation level and health Physics are also included. (Author)

  11. Detailed description of an SSAC at the facility level for on-load refueled power reactor facilities

    International Nuclear Information System (INIS)

    Jones, R.J.

    1985-11-01

    The purpose of this document is to provide a detailed description of a system for the accounting for and control of nuclear material in an on-load refueled power reactor facility which can be used by a facility operator to establish his own system to comply with a national system for nuclear material accounting and control and to facilitate application of IAEA safeguards. The scope of this document is limited to descriptions of the following SSAC elements: (1) Nuclear Material Measurements; (2) Measurement Quality; (3) Records and Reports; (4) Physical Inventory Taking; (5) Material Balance Closing

  12. Heat removing device for nuclear reactor container facility

    Energy Technology Data Exchange (ETDEWEB)

    Tateno, Seiya; Tominaga, Kenji; Iwata, Yasutaka; Kinoshita, Shoichiro; Niino, Tsuyoshi

    1994-09-30

    A pressure suppression chamber incorporating pool water is disposed inside of a reactor container for condensating steams released to a dry well upon occurrence of abnormality. A pool is disposed at the outer circumference of the pressure suppression chamber having a steel wall surface of the reactor container as a partition wall. The outer circumferential pool is in communication with ocean by way of a lower communication pipeline and an upper communication pipeline. During normal plant operation state, partitioning valves disposed respectively to the upper and lower communication pipelines are closed, so that the outer circumferential pool is kept empty. After occurrence loss of coolant accident, steams generated by after-heat of the reactor core are condensated by pool water of the pressure suppression chamber, and the temperature of water in the pressure suppression chamber is gradually elevated. During the process, the partition valves of the upper and lower communication pipelines are opened to introduce cold seawater to the outer circumferential pool. With such procedures, heat of the outer circumferential pool is released to the sea by natural convection of seawater, thereby enabling to remove residual heat without dynamic equipments. (I.N.).

  13. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    International Nuclear Information System (INIS)

    Salazar, M.D.

    1998-01-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel

  14. Development programs on decommissioning technology for reactors and fuel cycle facilities in Japan

    International Nuclear Information System (INIS)

    Fujiki, K.

    1992-01-01

    The Science and Technology Agency (STA) of Japan is promoting technology development for decommissioning of nuclear facilities by entrusting various research programs to concerned research organisations: JAERI, PNC and RANDEC, including first full scale reactor decommissioning of JPDR. According to the results of these programs, significant improvement on dismantling techniques, decontamination, measurement etc. has been achieved. Further development of advanced decommissioning technology has been started in order to achieve reduction of duration of decommissioning work and occupational exposures in consideration of the decommissioning of reactors and fuel cycle facilities. (author) 5 refs.; 7 figs.; 1 tab

  15. Physics design of fast reactor safety test facilities for in-pile experiments

    International Nuclear Information System (INIS)

    Travelli, A.; Matos, J.E.; Snelgrove, J.L.; Shaftman, D.H.; Tzanos, C.P.; Lam, S.K.; Pennington, E.M.; Woodruff, W.L.

    1976-01-01

    A determined effort to identify and resolve current Fast Breeder Reactor safety testing needs has recently resulted in a number of conceptual designs for FBR safety test facilities which are very complex and diverse both in their features and in their purpose. The paper discusses the physics foundations common to most fast reactor safety test facilities and the constraints which they impose on the design. The logical evolution, features, and capabilities of several major conceptual designs are discussed on the basis of this common background

  16. Los Alamos National Laboratory case studies on decommissioning of research reactors and a small nuclear facility

    Energy Technology Data Exchange (ETDEWEB)

    Salazar, M.D.

    1998-12-01

    Approximately 200 contaminated surplus structures require decommissioning at Los Alamos National Laboratory. During the last 10 years, 50 of these structures have undergone decommissioning. These facilities vary from experimental research reactors to process/research facilities contaminated with plutonium-enriched uranium, tritium, and high explosives. Three case studies are presented: (1) a filter building contaminated with transuranic radionuclides; (2) a historical water boiler that operated with a uranyl-nitrate solution; and (3) the ultra-high-temperature reactor experiment, which used enriched uranium as fuel.

  17. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  18. Calculation and experimental measurements in the Argonauta reactor subcritical and exponential facility

    International Nuclear Information System (INIS)

    Voi, Dante L.; Furieri, Rosane C.A.A.; Renke, Carlos A.C.; Bastos, Wilma S.; Ferreira, Francisco J.O.

    1997-01-01

    Initial measurements were performed on the exponential and subcritical facility installed on the internal thermal column of the Argonauta reactor at IEN-CNEN-Rio de Janeiro, Brazil. The measurements are include in the reactor physics experimental program for integral parameters determination, for both valid and confirmed theoretical models for reactor calculation. Gamma doses and neutron fluxes were measured with telescopic, proportional counters, wire and foil detectors. Experimental data were compared with results obtained by application of CITATION code. (author). 4 refs., 8 figs

  19. The FRJ 1 reactor (MERLIN) at Juelich, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FRJ 1 reactor and associated hot cell facilities, with the main emphasis on experimental irradiation facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  20. Nuclear facilities of EdF's operational hot base of Tricastin. 2009 annual report

    International Nuclear Information System (INIS)

    2010-01-01

    This annual report is established on account of article 21 of the 2006-686 French law from June 13, 2006, relative to the transparency and safety in the nuclear domain. It describes, first, the nuclear facilities of the EdF operational hot base of Tricastin, then, the measures taken to ensure their safety (personnel radioprotection, actions implemented for nuclear safety improvement, organisation in crisis situation, external and internal controls, technical assessment of the facilities), and finally the procedures of management of radioactive wastes. A glossary and the viewpoint of the Committee of Hygiene, safety and working conditions about the content of the document conclude the report. (J.S.)

  1. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    International Nuclear Information System (INIS)

    1997-01-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits

  2. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  3. Thermal hydraulic modelling of the Mo and Iridium irradiation facilities of the RA10 reactor

    International Nuclear Information System (INIS)

    Gramajo, M.; García, J.; Marcel, C.P.

    2013-01-01

    The RA-10 reactor is a multipurpose, open pool research reactor. The core consists of a rectangular array of MTR type fuel. The produced thermal power is 30 MW which is extracted by the refrigeration system via an ascendant flow through the core. The core reflector is D 2 O contained in a watertight tank. The design of the reactor includes a number of out-core facilities which are meant to be used for industrial, medical and research purposes. Among all the facilities, the most important ones are the Molybdenum and Iridium ones which we modeled in this work. During the normal operation of the reactor, the manipulation and the on-line extraction of the irradiation facilities is foreseen. Therefore the study of the head loss during the normal operation as well as during the extraction maneuvers plays a relevant role in the design and safety analysis. In this work a CFD commercial code is use dto perform the calculations needed to guarantee the design requirements.In addition, a full detailed geometric model for both, the Molybdenum and Iridium facilities,is used to perform the required simulations. The obtained results allow to evaluating the thermal-hydraulic performance of the proposed facilities designs. (author)

  4. Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

    International Nuclear Information System (INIS)

    1982-09-01

    The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

  5. Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1983-10-01

    The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

  6. Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-09-01

    The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

  7. Fourteenth Annual Meeting of the International Working Group on Past Reactors. Summary Report. Part I

    International Nuclear Information System (INIS)

    1981-11-01

    The Fourteenth Annual Meeting of the IAEA-IWGFR was held in accordance with the recommendations of the previous Annual Group Meeting, at the Vienna International Centre, Vienna from 31 March to 3 April 1981. All Member States of the group were represented at the meeting: France, the Federal Republic of Germany, India, Italy, Japan, the Union of Soviet Socialist Republics, the United Kingdom and the United States of America. The meeting was also attended by representatives from the Commission of European Communities, the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development, the International Atomic Energy Agency and observers from Switzerland. The Agenda of the Meeting was as follows: 1. Review of IWGFR activities; 2. Consideration of future method of operation of the IWGFR; 3. Consideration of Conferences on Fast Reactors; 4. Consideration of the major recommendations of some of the IWGFR specialists' meetings for which the support of the IWGFR is requested; 5. Consideration of a schedule for specialists' meetings in 1981-1982; 6. Presentations and discussions on national programmes on fast breeder reactors.; 7. Recommendation of the IWGFR regarding a request of Switzerland concerning participation in the IWGFR; 8. The date and place of the Fifteenth Annual Meeting of the IWGFR

  8. Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

    International Nuclear Information System (INIS)

    1983-10-01

    The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

  9. Minthorn Springs Creek summer juvenile release and adult collection facility: Annual report 1992; ANNUAL

    International Nuclear Information System (INIS)

    Rowan, Gerald D.

    1993-01-01

    The Confederated Tribes of the Umatilla Indian Reservation (CT'UIR) and Oregon Department of Fish and Wildlife (ODFW) are cooperating in a joint effort to supplement steelhead and re-establish salmon runs in the Umatilla River Basin. As an integral part of this program, Bonifer and Minthorn Acclimation Facilities are operated for holding and spawning adult steelhead and fall chinook salmon and acclimation and release of juvenile salmon and steelhead. Acclimation of 109,101 spring chinook salmon and 19,977 summer steelhead was completed at Bonifer in the spring of 1992. At Minthorn, 47,458 summer steelhead were acclimated and released. Control groups of spring chinook salmon were released instream concurrent with the acclimated releases to evaluate the effects of acclimation on adult returns to the Umatilla River. Acclimation studies with summer steelhead were not conducted in 1992. A total of 237 unmarked adult steelhead were collected for broodstock at Three Mile Dam from October 18, 1991 through April 24, 1992 and held at Minthorn. Utilizing a 3 x 3 spawning matrix, a total of 476,871 green eggs were taken from 86 females. The eggs were transferred to Umatilla Hatchery for incubation, rearing, and later release into the Umatilla River. A total of 211 fall chinook salmon were also collected for broodstock at Three Mile Dam and held at Minthorn. Using a 1:1 spawning ratio, a total of 195,637 green eggs were taken from 58 females. They were also transferred to Umatilla Hatchery for incubation, rearing, and later release into the Umatilla River. Personnel from the ODFW Eastern Oregon Fish Pathology Laboratory in La Grande took samples of tissues and reproductive fluids from Umatilla River summer steelhead and fall chinook salmon broodstock for monitoring and evaluation purposes. Cell culture assays for replicating agents, including IHNV virus, on all spawned fish were negative. One of 60 summer steelhead tested positive for EIBS virus, while all fall chinook tested

  10. Good Practices for Water Quality Management in Research Reactors and Spent Fuel Storage Facilities

    International Nuclear Information System (INIS)

    2011-01-01

    Water is the most common fluid used to remove the heat produced in a research reactor (RR). It is also the most common media used to store spent fuel elements after being removed from the reactor core. Spent fuel is stored either in the at-reactor pool or in away-from-reactor wet facilities, where the fuel elements are maintained until submission to final disposal, or until the decay heat is low enough to allow migration to a dry storage facility. Maintaining high quality water is the most important factor in preventing degradation of aluminium clad fuel elements, and other structural components in water cooled research reactors. Excellent water quality in spent fuel wet storage facilities is essential to achieve optimum storage performance. Experience shows the remarkable success of many research reactors where the water chemistry has been well controlled. In these cases, aluminium clad fuel elements and aluminium pool liners show few, if any, signs of either localized or general corrosion, even after more than 30 years of exposure to research reactor water. In contrast, when water quality was allowed to degrade, the fuel clad and the structural parts of the reactor have been seriously corroded. The driving force to prepare this publication was the recognition that, even though a great deal of information on research reactor water quality is available in the open literature, no comprehensive report addressing the rationale of water quality management in research reactors has been published to date. This report is designed to provide a comprehensive catalogue of good practices for the management of water quality in research reactors. It also presents a brief description of the corrosion process that affects the components of a research reactor. Further, the report provides a basic understanding of water chemistry and its influence on the corrosion process; specifies requirements and operational limits for water purification systems of RRs; describes good practices

  11. The advanced neutron source - A world-class research reactor facility

    International Nuclear Information System (INIS)

    Thompson, P.B.; Meek, W.E.

    1993-01-01

    The advanced neutron source (ANS) is a new facility being designed at the Oak Ridge National Laboratory that is based on a heavy-water-moderated reactor and extensive experiment and user-support facilities. The primary purpose of the ANS is to provide world-class facilities for neutron scattering research, isotope production, and materials irradiation in the United States. The neutrons provided by the reactor will be thermalized to produce sources of hot, thermal, cold, very cold, and ultracold neutrons usable at the experiment stations. Beams of cold neutrons will be directed into a large guide hall using neutron guide technology, greatly enhancing the number of research stations possible in the project. Fundamental and nuclear physics, materials analysis, and other research pro- grams will share the neutron beam facilities. Sufficient laboratory and office space will be provided to create an effective user-oriented environment

  12. 2016 Annual Reuse Report for the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-02-01

    This report describes conditions and information, as required by the state of Idaho, Department of Environmental Quality Reuse Permit I-161-02, for the Advanced Test Reactor Complex Cold Waste Ponds located at Idaho National Laboratory from November 1, 2015–October 31, 2016. The effective date of Reuse Permit I-161-02 is November 20, 2014 with an expiration date of November 19, 2019. This report contains the following information: • Facility and system description • Permit required effluent monitoring data and loading rates • Permit required groundwater monitoring data • Status of compliance activities • Issues • Discussion of the facility’s environmental impacts. During the 2016 permit year, 180.99 million gallons of wastewater were discharged to the Cold Waste Ponds. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest in well USGS-065, which is the closest downgradient well to the Cold Waste Ponds. Sulfate and total dissolved solids concentrations decrease rapidly as the distance downgradient from the Cold Waste Ponds increases. Although concentrations of sulfate and total dissolved solids are significantly higher in well USGS-065 than in the other monitoring wells, both parameters remained below the Ground Water Quality Rule Secondary Constituent Standards in well USGS-065. The facility was in compliance with the Reuse Permit during the 2016 permit year.

  13. 2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond

    International Nuclear Information System (INIS)

    Lewis, Mike

    2012-01-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA 000161 01, Modification B), for the wastewater land application site at the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: Facility and system description Permit required effluent monitoring data and loading rates Groundwater monitoring data Status of compliance activities Noncompliance and other issues Discussion of the facility's environmental impacts During the 2011 permit year, approximately 166 million gallons of wastewater were discharged to the Cold Waste Pond. This is well below the maximum annual permit limit of 375 million gallons. As shown by the groundwater sampling data, sulfate and total dissolved solids concentrations are highest near the Cold Waste Pond and decrease rapidly as the distance from the Cold Waste Pond increases. Although concentrations of sulfate and total dissolved solids are elevated near the Cold Waste Pond, both parameters were below the Ground Water Quality Rule Secondary Constituent Standards in the down gradient monitoring wells.

  14. Upgrading of neutron radiography/tomography facility at research reactor

    International Nuclear Information System (INIS)

    Abd El Bar, Waleed; Mongy, Tarek

    2014-01-01

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  15. Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford site

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1989-01-01

    This paper presents the methods used during the walkdowns to compile as-built structural information to seismically qualify or verify the seismic adequacy of safety class components in the Plutonium Finishing Plant complex. The Plutonium finishing Plant is a non-reactor nuclear facility built during the 1950's and was designed to the Uniform Building Code criteria for both seismic and wind events. This facility is located at the US Department of Energy Hanford Site near Richland, Washington

  16. Status and future program of reactor physics experiments in JAERI Critical facilities, FCA and TCA

    International Nuclear Information System (INIS)

    Okajima, Shigeaki; Osugi, Toshitaka; Nakajima, Ken; Suzaki, Takenori; Miyoshi, Yoshinori

    1999-01-01

    The critical facilities in JAERI, FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly), have been used to provide integral data for evaluation of nuclear data as well as for development of various types of reactor since they went critical in 1960's. In this paper a review is presented on the experimental programs in both facilities. And the experimental programs in next 5 years are also shown. (author)

  17. Operational experiences in radiation protection in fast reactor fuel reprocessing facility

    International Nuclear Information System (INIS)

    Meenakshisundaram, V.; Rajagopal, V.; Santhanam, R.; Baskar, S.; Madhusoodanan, U.; Chandrasekaran, S.; Balasundar, S.; Suresh, K.; Ajoy, K.C.; Dhanasekaran, A.; Akila, R.; Indira, R.

    2008-01-01

    The Compact Reprocessing facility for Advanced fuels in Lead cells (CORAL), situated at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam is a pilot plant to reprocess the mixed carbide fuel, for the first time in the world. Reprocessing of fuel with varying burn-ups up to 155 G Wd/t, irradiated at Fast Breeder Test Reactor (FBTR), has been successfully carried out at CORAL. Providing radiological surveillance in a fuel reprocessing facility itself is a challenging task, considering the dynamic status of the sources and the proximity of the operator with the radioactive material and it is more so in a fast reactor fuel reprocessing facility due to handling of higher burn-up fuels associated with radiation fields and elevated levels of fissile material content from the point of view of criticality hazard. A very detailed radiation protection program is in place at CORAL. This includes, among others, monitoring the release of 85 Kr and other fission products and actinides, if any, through stack on a continuous basis to comply with the regulatory limits and management of disposal of different types of radioactive wastes. Providing radiological surveillance during the operations such as fuel transport, chopping and dissolution and extraction cycle was without any major difficulty, as these were carried out in well-shielded and high integrity lead cells. Enforcement of exposure control assumes more importance during the analysis of process samples and re-conversion operations due to the presence of fission product impurities and also since the operations were done in glove boxes and fume hoods. Although the radiation fields encountered in process area were marginally higher, due to the enforcement of strict administrative controls, the annual exposure to the radiation workers was well within the regulatory limit. As the facility is being used as test bed for validation of prototype equipment, periodic inspection and maintenance of components such as centrifuge

  18. Abbreviated sampling and analysis plan for planning decontamination and decommissioning at Test Reactor Area (TRA) facilities

    International Nuclear Information System (INIS)

    1994-10-01

    The objective is to sample and analyze for the presence of gamma emitting isotopes and hazardous constituents within certain areas of the Test Reactor Area (TRA), prior to D and D activities. The TRA is composed of three major reactor facilities and three smaller reactors built in support of programs studying the performance of reactor materials and components under high neutron flux conditions. The Materials Testing Reactor (MTR) and Engineering Test Reactor (ETR) facilities are currently pending D/D. Work consists of pre-D and D sampling of designated TRA (primarily ETR) process areas. This report addresses only a limited subset of the samples which will eventually be required to characterize MTR and ETR and plan their D and D. Sampling which is addressed in this document is intended to support planned D and D work which is funded at the present time. Biased samples, based on process knowledge and plant configuration, are to be performed. The multiple process areas which may be potentially sampled will be initially characterized by obtaining data for upstream source areas which, based on facility configuration, would affect downstream and as yet unsampled, process areas. Sampling and analysis will be conducted to determine the level of gamma emitting isotopes and hazardous constituents present in designated areas within buildings TRA-612, 642, 643, 644, 645, 647, 648, 663; and in the soils surrounding Facility TRA-611. These data will be used to plan the D and D and help determine disposition of material by D and D personnel. Both MTR and ETR facilities will eventually be decommissioned by total dismantlement so that the area can be restored to its original condition

  19. Research nuclear reactor RA - Annual Report 1997; Istrazivacki nuklearni reaktor RA - Izvestaj za 1997. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Vinca Institute of Nuclear Sciences, Beograd (Serbia and Montenegro)

    1997-12-01

    RA reactor is not in operation since 1984, activities related to revitalisation of the RA reactor started in 1986. The planned actions related to renewal of the reactor components were finished except for the most important action, related to exchange of complete reactor instrumentation which was delayed. Only 80% of the instrumentation was delivered until September 1991. Since then any delivery of components to Yugoslavia was stopped because of the sanctions imposed to our country. The existing RA reactor instrumentation was dismantled. Control and maintenance of the reactor components was done regularly and efficiently. Fuel inspection by the IAEA safeguards inspectors was done on a monthly basis. There have been on the average 42 employees at the RA reactor which is considered sufficient for maintenance and repair conditions. The problem of financing the reactor activities and maintenance remains unsolved. Research reactor RA Annual report for year 1997 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. [Serbo-Croat] Reaktor RA nije u pogonu od 1984, aktivnosti na revitalizaciji, rekostrukciji i modernizaciji reaktorski sistema zapocete su 1986. godine. Okoncan je niz zahvata na opremi postrojenja kojima ce se u narednom periodu omoguciti kontinualan i pouzdan rad ovog reaktora. Poslednji, i ujedno najveci zahvat, koji se odnosi na zamenu celokupne instrumentacije kasni zbog zastoja u isporuci opreme koja se izradjuje u Sovjetskom savezu. Do septembra 1991. godine isporuceno svega 80% od predvidjene kolicine. Od tada je svaka isporuka obustavljena, a razlog je privremena zabrana na sve isporuke opreme za Jugoslaviju usled sankcija uvedenih od strane organizacije Ujedinjenih nacija. Demontirana je postojeca instrumentacija. Kontrola i odrzavanje celopkupne opreme postrojenja, kao i remontni radovi izvrsavani su redovno i efikasno. Kontrola goriva od strane safeguard inspektora MAAE obavljana

  20. Aqueous Boric acid injection facility of PWR type reactor

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi; Iwami, Masao.

    1996-01-01

    If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

  1. Research nuclear reactor RA - Annual report 1992; Istrazivacki nuklearni reaktor RA - Izvestaj za 1992. godinu

    Energy Technology Data Exchange (ETDEWEB)

    Sotic, O [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1992-12-01

    Research reactor RA Annual report for year 1992 is divided into two main parts to cover: (1) operation and maintenance and (2) activities related to radiation protection. First part includes 8 annexes describing reactor operation, activities of services for maintenance of reactor components and instrumentation, financial report and staffing. Second annex B is a paper by Z. Vukadin 'Recurrence formulas for evaluating expansion series of depletion functions' published in 'Kerntechnik' 56, (1991) No.6 (INIS record no. 23024136. Second part of the report is devoted to radiation protection issues and contains 4 annexes with data about radiation control of the working environment and reactor environment, description of decontamination activities, collection of radioactive wastes, and meteorology data. [Serbo-Croat] Godisnji izvestaj o radu nuklearnog reaktora RA za 1992. godinu sastoji se od dva dela: prvi deo obuhvata pogon i odzavanje reaktora RA, a drugi poslove zastite od zracenja na reaktoru RA. Prvi deo sadrzi 8 priloga, koji opisuju rad reaktora i poslove sluzbi za odrzavanje opreme i komponenti, finansijski izvestaj, kadrovsku strukturu osoblja reaktora. Drugi prilog (B) je rad Z. Vukadina 'Recurrence formulas for evaluating expansion series of depletion functions' objavljen u casopisu Kerntechnik, 1991. Drugi deo izvestaja o poslovima zastite od zracenja sadrzi 4 priloga sa podacima radijacione kontrole radne sredine i okoline reaktora, opis poslova dekontaminacije i sakupljanja radioaktivnih materija, kao i meteoroloske podatke.

  2. Technical feasibility of an Integral Fast Reactor (IFR) as a future option for fast reactor cycles. Integrate a small metal-fueled fast reactor and pyroprocessing facilities

    International Nuclear Information System (INIS)

    Tanaka, Nobuo

    2017-01-01

    Integral Fast Reactor that integrated fast reactor and pyrorocessing facilities developed by Argonne National Laboratory in the U.S. is an excellent nuclear fuel cycle system for passive safety, nuclear non-proliferation, and reduction in radioactive waste. In addition, this system can be considered as a technology applicable to the treatment of the fuel debris caused by the Fukushima Daiichi Nuclear Power Station accident. This study assessed the time required for debris processing, safety of the facilities, and construction cost when using this technology, and examined technological possibility including future technological issues. In a small metal-fueled reactor, it is important to design the core that achieves both of reduction in combustion reactivity and reduction in coolant reactivity. In system design, calorimetric analysis, structure soundness assessment, seismic feasibility establishment study, etc. are important. Regarding safety, research and testing are necessary on the capabilities of passive reactor shutdown and reactor core cooling as well as measures for avoiding re-criticality, even when emergency stop has failed. In dry reprocessing system, studies on electrolytic reduction and electrolytic refining process for treating the debris with compositions different from those of normal fuel are necessary. (A.O.)

  3. Scottish Universities Research and Reactor Centre annual report 1987-1988

    International Nuclear Information System (INIS)

    Whitley, J.E.

    1988-01-01

    The Scottish Universities Research and Reactor Centre (SURRC) provides facilities for research in isotopic, nuclear and earth sciences and collaborates with Scottish University departments on a wide range of research topics. One of its main areas of work is the Isotope Geology Unit. This has worked with the Nuclear Medicine Unit on the application of enriched stable isotope tracers in the biological and clinical sciences. The measurement of radioactive isomers is applied to quaternary geology, archaeology, nuclear medicine, health physics, oceanography, atomospheric sciences, environmental chemistry, nuclear waste disposal and mathematical modelling of the environment. There are also radiocarbon dating facilities. The facilities and the research undertaken at the Centre in the year 1987-1988, the Centre's twenty-fifth year are summarized in this report. (U.K.)

  4. A complete fuel development facility utilizing a dual core TRIGA reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, A; Law, G C [General Atomic Co., San Diego, CA (United States)

    1974-07-01

    A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 10{sup 14} n/cm{sup 2}-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 10{sup 17} n/cm{sup 2}-sec. The pulse width at

  5. Research nuclear reactor RA - Annual Report 1990 with the comparative evaluation for the period 1986-1990

    International Nuclear Information System (INIS)

    Sotic, O.

    1990-12-01

    Annual report concerning the project 'RA research nuclear reactor' for 1990, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities [sr

  6. Seed irradiation facilities at TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Najzer, M.

    1972-01-01

    Fast neutrons and gamma-rays with their high and low LET respectively are excellent complementary tools for investigation of the effect of different types of mutations. TRIGA Irradiation Facility and Thermal Column Irradiation Facility were designed and installed for the first time in the TRIGA tank and thermal column respectively. The basic idea of design was the use of depleted uranium as gamma-ray and thermal neutron shield and simultaneously as thermal to fast neutron converter. Low LET radiation, due to direct and thermal neutron capture gamma-rays, is strongly attenuated while fast neutron flux is increased. GIF is made of a cadmium tube inserted in a graphite block. It is located in the central thermal column channel. The basic idea is to convert thermal neutrons to gamma-rays by capture in the cadmium

  7. [Fortieth Annual] Meeting of the Technical Working Group on Fast Reactors (TWG-FR). Working Material

    International Nuclear Information System (INIS)

    2008-01-01

    The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 39th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2007, 2008 and 2009; - Review the Agency’s ongoing information exchange and co-ordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations; - Discuss future joint activities in view of the Agency’s Programme and Budget Cycle 2008–2009 (and beyond)

  8. Education and training at the Rensselaer Polytechnic Institute reactor critical facility

    International Nuclear Information System (INIS)

    Harris, D.R.

    1989-01-01

    The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

  9. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1982-August 1983

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1984-03-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1982 was over 5500 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1982 through August 1983

  10. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1983-August 1984

    International Nuclear Information System (INIS)

    Mulder, R.U.; Benneche, P.E.

    1984-11-01

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1983 was over 6000 megawatt-hours) research reactor in the mid-Atlantic states. In addition, a second, small (50 watt) reactor is also available for use in educational and research programs. A major objective of this facility is to expand its support of educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1983 through August 1984

  11. PANDA a multi-purpose thermal-hydraulics facility devoted to nuclear reactor containment safety analysis

    International Nuclear Information System (INIS)

    Paladino, Domenico

    2014-01-01

    This paper presents the multi purpose facility PANDA devised for the safety analysis of nuclear reactor containment. The passive safety systems for LWRs have been explained with details about the PAssive Nachzerfallswärmeabfuhr und Druck-Abbau Testanlage (PANDA)

  12. Analyses in support of the Laboratory Microfusion Facility and ICF commercial reactor designs

    International Nuclear Information System (INIS)

    Meier, W.R.; Monsler, M.J.

    1988-01-01

    Our work on this contract was divided into two major categories; two thirds of the total effort was in support of the Laboratory Microfusion Facility (LMF), and one third of the effort was in support of Inertial Confinement Fusion (ICF) commercial reactors. This final report includes copies of the formal reports, memoranda, and viewgraph presentations that were completed under this contract

  13. Test and application of thermal neutron radiography facility at Xi'an pulsed reactor

    CERN Document Server

    Yang Jun; Zhao Xiang Feng; Wang Dao Hua

    2002-01-01

    A thermal neutron radiography facility at Xi'an Pulsed Reactor is described as well as its characteristics and application. The experiment results show the inherent unsharpness of BAS ND is 0.15 mm. The efficient thermal neutron n/gamma ratio is lower in not only steady state configuration but also pulsing state configuration and it is improved using Pb filter

  14. The Text of the Agreement for the Application of Agency Safeguards to United States Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1964-08-14

    The text of the Agreement between the Agency and the Government of the United States of America for the application of Agency safeguards to United States reactor facilities, which was signed on 15 June 1964 and entered into force on 1 August 1964, is reproduced in this document for the information of all Members.

  15. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24

  16. An automatic device for sample insertion and extraction to/from reactor irradiation facilities

    International Nuclear Information System (INIS)

    Alloni, L.; Venturelli, A.; Meloni, S.

    1990-01-01

    At the previous European Triga Users Conference in Vienna,a paper was given describing a new handling tool for irradiated samples at the L.E.N.A plant. This tool was the first part of an automatic device for the management of samples to be irradiated in the TRIGA MARK ii reactor and successively extracted and stored. So far sample insertion and extraction to/from irradiation facilities available on reactor top (central thimble,rotatory specimen rack and channel f),has been carried out manually by reactor and health-physics operators using the ''traditional'' fishing pole provided by General Atomic, thus exposing reactor personnel to ''unjustified'' radiation doses. The present paper describes the design and the operation of a new device, a ''robot''type machine,which, remotely operated, takes care of sample insertion into the different irradiation facilities,sample extraction after irradiation and connection to the storage pits already described. The extraction of irradiated sample does not require the presence of reactor personnel on the reactor top and,therefore,radiation doses are strongly reduced. All work from design to construction has been carried out by the personnel of the electronic group of the L.E.N.A plant. (orig.)

  17. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  18. University of Florida Training Reactor: Annual progress report, September 1, 1986-August 31, 1987

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1987-11-01

    The University of Florida Training Reactor's overall utilization for the past reporting year (September 1986 through August 1987) has returned to the increased levels of quality usage characteristic of the two years prior to the last reporting year when the maintenance outage to repair sticking control blades reduced availability for the year to near 50%. Indeed, the 91.5% availability factor for this reporting year is the highest in the last five years and probably in the 27 year history of the facility. As a final statement on the effectiveness of the corrective maintenance last year, the overall availability factor has been over 94% since returning to normal operations. The UFTR continues to experience a high rate of utilization in a broad spectrum of areas with total utilization continuing near the highest levels recorded in the early 1970's. This increase has been supported by a variety of usages ranging from research and educational utilization by users within the University of Florida as well as other researchers and educators around the state of Florida through the support of the DOE Reactor Sharing Program and several externally supported usages. Significant usage has also been devoted to facility enhancement where a key ingredient for this usage has been a stable management staff. Uses, reactor operation, maintenance, technical specifications, radioactive releases, and research programs are described in this report

  19. A study on the direct use of spent PWR fuel in CANDU reactors. DUPIC facility engineering

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyun Soo; Lee, Jae Sul; Choi, Jong Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This report summarizes the second year progress of phase II of DUPIC program which aims to verify experimentally the feasibility of direct use of spent PWR fuel in CANDU reactors. The project is to provide the experimental facilities and technologies that are required to perform the DUPIC experiment. As an early part of the project, engineering analysis of those facilities and construction of mock-up facility are described. Another scope of the project is to assess the DUPIC fuel cycle system and facilitate international cooperation. The progresses in this scope of work made during the fiscal year are also summarized in the report. 38 figs, 44 tabs, 8 refs. (Author).

  20. Operator training facilities for CEGB advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Green, J.F.; Birnie, S.

    1980-01-01

    The facilities provided at the Nuclear Power Training Centre of the CEGB for the training of operators fo the AGR are described. The simulator control desks are replicas of three AGR designs with, in addition, simulation of the Data Processing System for each station. Three modes of operation are envisaged: a.) Demonstration where the simulator is used by the tutor to illustrate lecture on plant behaviour. b.) Interaction where the student carries out normal procedures and experiences plant failure situations. c.) Investigation where engineering staff use the simulator for validation of modified operational procedures, ergonomic studies etc. (orig./HP)

  1. LASTRON - Second generation accelerators and chemical reactors for EBFGT facilities

    International Nuclear Information System (INIS)

    Edinger, R.

    2011-01-01

    Commercializing reliable affordable electron beam flue gas treatment technology requires both, the optimization of accelerator technology and chemical reaction chambers. Moreover, this engineering process involves the integration of beam specific characteristics, such as dosage distribution and penetration of electrons into the flue gas stream. In consideration of the treatment economy, it might be required to calculate the overall process performance without merely limiting the evaluation to accelerator efficiency. For example, a higher energy beam, 1MeV to 2 MeV, reduces the losses in the beam window and penetrates further into the gas stream and, therefore, increases the overall process economy. The energy distribution should be optimized with respect to the configuration of the chemical reaction chamber in order to treat the flue gas uniformly. All these measures are required to achieve high removal rates in large flue gas streams. Today removal rates of more than 99% SO x and more than 80% SO x are required to be compliant with future emission legislations. It is planed to establish a 100,000m³ electron beam flue gas treatment facility that can achieve constant removal rates of higher than 99.4% SO x and more than 80% NO x . The high removal rates would allow us to place CO 2 capture technologies down stream of the EBFGT facility. (author)

  2. LASTRON - Second generation accelerators and chemical reactors for EBFGT facilities

    Energy Technology Data Exchange (ETDEWEB)

    Edinger, R. [PAVAC Industries Inc. Richmond, British Columbia (Canada)

    2011-07-01

    Commercializing reliable affordable electron beam flue gas treatment technology requires both, the optimization of accelerator technology and chemical reaction chambers. Moreover, this engineering process involves the integration of beam specific characteristics, such as dosage distribution and penetration of electrons into the flue gas stream. In consideration of the treatment economy, it might be required to calculate the overall process performance without merely limiting the evaluation to accelerator efficiency. For example, a higher energy beam, 1MeV to 2 MeV, reduces the losses in the beam window and penetrates further into the gas stream and, therefore, increases the overall process economy. The energy distribution should be optimized with respect to the configuration of the chemical reaction chamber in order to treat the flue gas uniformly. All these measures are required to achieve high removal rates in large flue gas streams. Today removal rates of more than 99% SO{sub x} and more than 80% SO{sub x} are required to be compliant with future emission legislations. It is planed to establish a 100,000m³ electron beam flue gas treatment facility that can achieve constant removal rates of higher than 99.4% SO{sub x} and more than 80% NO{sub x}. The high removal rates would allow us to place CO{sub 2} capture technologies down stream of the EBFGT facility. (author)

  3. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  4. Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

    International Nuclear Information System (INIS)

    Oakley, D.J.

    1978-01-01

    The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

  5. Advancing nuclear technology and research. The advanced test reactor national scientific user facility

    Energy Technology Data Exchange (ETDEWEB)

    Benson, Jeff B; Marshall, Frances M [Idaho National Laboratory, Idaho Falls, ID (United States); Allen, Todd R [Univ. of Wisconsin, Madison, WI (United States)

    2012-03-15

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research. The mission of the ATR NSUF is to provide access to world-class facilities, thereby facilitating the advancement of nuclear science and technology. Cost free access to the ATR, INL post irradiation examination facilities, and partner facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to United States Department of Energy. To increase overall research capability, ATR NSUF seeks to form strategic partnerships with university facilities that add significant nuclear research capability to the ATR NSUF and are accessible to all ATR NSUF users. (author)

  6. Final report of the HFIR [High Flux Isotope Reactor] irradiation facilities improvement project

    International Nuclear Information System (INIS)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987

  7. Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Ishihara, Noriyuki

    1985-01-01

    The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

  8. The FR 2 reactor at Karlsruhe, F.R. Germany and associated hot cell facilities. Information sheets

    International Nuclear Information System (INIS)

    Hardt, P. von der; Roettger, H.

    1981-01-01

    Technical information is given on the FR 2 reactor and associated hot cell facilities, specialized irradiation devices (loops and capsules) and possibilities for post-irradiation examinations of samples. The information is presented in the form of eight information sheets under the headings: main characteristics of the reactor; utilization and specialization of the reactor; experimental facilities; neutron spectra; main characteristics of specialized irradiation devices; main characteristics of hot cell facilities; equipment and techniques available for post-irradiation examinations; utilization and specialization of the hot cell facilities

  9. In-pile experiments and test facilities proposed for fast reactor safety

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Avery, R.; Goldman, A.J.; Fauske, H.K.; Marchaterre, J.F.; Rose, D.; Wright, A.E.

    1976-01-01

    The role of in-pile experiments in support of the resolution of fast breeder reactor safety and licensing issues has been re-examined, with emphasis on key safety issues. Experiment needs have been related to the specific characteristics of these safety issues and to realistic requirements for additional test facility capabilities which can be achieved and utilized within the next ten years. It is found that those safety issues related to the energetics of core disruptive accidents have the largest impact on new facility requirements. However, utilization of existing facilities with modifications can provide for a continuing increase in experiment capability and experiment results on a timely bases. Emphasis has been placed upon maximum utilization of existing facilities and minimum requirements for new facilities. This evaluation has concluded that a new Safety Test Facility, STF, along with major modifications to the EBR II facility, improvement in TREAT capabilities, the existing Sodium Loop Safety Facility and corresponding Support Facilities provide the essential elements of the Safety Research Experiment Facilities (SAREF) required for resolution of key issues

  10. The Advanced Neutron Source (ANS) project: A world-class research reactor facility

    International Nuclear Information System (INIS)

    Thompson, P.B.; Meek, W.E.

    1993-01-01

    This paper provides an overview of the Advanced Neutron Source (ANS), a new research facility being designed at Oak Ridge National Laboratory. The facility is based on a 330 MW, heavy-water cooled and reflected reactor as the neutron source, with a thermal neutron flux of about 7.5x10 19 m -2 ·sec -1 . Within the reflector region will be one hot source which will serve 2 hot neutron beam tubes, two cryogenic cold sources serving fourteen cold neutron beam tubes, two very cold beam tubes, and seven thermal neutron beam tubes. In addition there will be ten positions for materials irradiation experiments, five of them instrumented. The paper touches on the project status, safety concerns, cost estimates and scheduling, a description of the site, the reactor, and the arrangements of the facilities

  11. Description of the PIE facility for research reactors irradiated fuels in CNEA

    International Nuclear Information System (INIS)

    Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

    2002-01-01

    The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

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

    International Nuclear Information System (INIS)

    Tegas Sutondo

    2012-01-01

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

  13. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

    1983-01-01

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

  14. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  15. Fast-neutron dosimetry in the seed-irradiation facility, ASTRA reactor, Seibersdorf

    International Nuclear Information System (INIS)

    Ahnstroem, G.; Burtscher, A.; Casta, J.

    1967-01-01

    An important part of the co-ordinated programme on the neutron irradiation of seeds has been the construction of a fast-neutron irradiation facility for swimming-pool reactors. This facility was installed around 70 cm from the core in the ASTRA reactor swimming-pool at the end of December, 1966. Also, for this programme a pair of constant potential ionization chambers have been constructed at the Institute of Biochemistry, Stockholm University. These chambes are of the type described in the technical annex and are the same size as the seed-irradiation vials to be used in the seed-irradiation container (diam. =15 mm, length = 60 mm). Some preliminary dosimetry experiments were undertaken to test the irradiation facility and the ionization chambers, and to investigate the usefulness of the dosimetry instructions in the Technical Annex. The results of these experiments are discussed in this paper. 3 refs, 6 figs, 7 tabs

  16. UNC Nuclear Industries reactor and fuels production facilities. 1984 effluent release report

    International Nuclear Information System (INIS)

    Rokkan, D.J.

    1985-01-01

    This document has been prepared to fulfill the annual reporting requirements of DOE 5484.1, ''Environmental Protection, Safety, and Health Protection Information Reporting Requirements.'' Radioanalyses performed on routine samples of liquid and airborne streams were evaluated using UNC's Environmental Release Summary computer program. All identified significant discharges from UNC facilities to the environment during CY 1984 are reported in this document

  17. Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part I

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1979-05-01

    The Twelfth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous AGM,in Vienna from 27 to 30 March 1979. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

  18. Twelfth annual meeting of the International Working Group on Fast Reactors. Summary report. Part I

    International Nuclear Information System (INIS)

    1979-05-01

    The Twelfth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous AGM,in Vienna from 27 to 30 March 1979. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

  19. Annual progress report of the University of Florida Training Reactor, September 1, 1981-August 31, 1982

    International Nuclear Information System (INIS)

    Diaz, N.J.; Vernetson, W.G.

    1982-11-01

    The University of Florida Training Reactor's overall utilization for the past reporting year has decreased by about 50% compared to the previous year, approaching the low levels of utilization characteristic of the previous two reporting years ending in August 1979 and August 1980 respectively. The energy generation also continues to be far below average historical levels and represents a drop of nearly 50% from the improved level of the previous year. The UFTR continues to operate with an outstanding safety record and in full compliance with regulatory requirements. The reactor and associated facilities continue to maintain a high in-state visibility and strong industry relationship. It is hoped that more indirect industry training will be accomplished in the upcoming year

  20. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  1. The University of Missouri Research Reactor facility can melter system

    International Nuclear Information System (INIS)

    Edwards, C.B. Jr.; Olson, O.L.; Stevens, R.; Brugger, R.M.

    1987-01-01

    At the University of Missouri Research Reactor (MURR), a waste compacting system for reducing the volume of radioactive aluminum cans has been designed, built and put into operation. In MURR's programs of producing radioisotopes and transmutation doping of silicon, a large volume of radioactive aluminum cans is generated. The Can Melter System (CMS) consists of a sorting station, a can masher, an electric furnace and a gas fired furnace. This system reduces the cans and other radioactive metal into barrels of solid metal close to theoretical density. The CMS has been in operation at the MURR now for over two years. Twelve hundred cu ft of cans and other metals have been reduced into 150 cu ft of shipable waste. The construction cost of the CMS was $4950.84 plus 1680 man hours of labor, and the operating cost of the CMS is $18/lb. The radiation exposure to the operator is 8.6 mR/cu ft. The yearly operating savings is $30,000. 20 figs., 10 tabs

  2. Guide to the safety design examination about light water reactor facilities for power generation

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    This guide was compiled to evaluate the validity of the design policy when the safety design is examined at the time of the application for approval of the installation of nuclear reactors. About 7 years has elapsed since the existing guide was established, and the more appropriate guide to evaluate the safety should be made on the basis of the knowledge and experience accumulated thereafter. The range of application of this guide is limited to the above described evaluation, and it is not intended as the general standard for the design of nuclear reactors. First, the definition of the words used in this guide is given. Then, the guide to the safety examination is described about the general matters of reactor facilities, nuclear reactors and the measuring and controlling system, reactor-stopping system, reactivity-controlling system and safety protection system, reactor-cooling system, reactor containment vessels, fuel handling and waste treatment system. Several matters which require attention in the application of this guide or the clarification of the significance and interpretation of the guide itself were found, therefore the explanation about them was added at the end of this guide. (Kako, I.)

  3. A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Hiroaki; An, Shigehiro.

    1983-01-01

    Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

  4. RIA testing capability of the transient reactor test facility

    International Nuclear Information System (INIS)

    Crawford, D.C.; Swanson, R.W.

    1999-01-01

    The advent of high-burnup fuel implementation in LWRs has generated international interest in high-burnup LWR fuel performance. Recent testing under simulated RIA conditions has demonstrated that certain fuel designs fail at peak fuel enthalpy values that are below existing regulatory criteria. Because many of these tests were performed with non-prototypically aggressive test conditions (i.e., with power pulse widths less than 10 msec FWHM and with non-protoypic coolant configurations), the results (although very informative) do not indisputably identify failure thresholds and fuel behavior. The capability of the TREAT facility to perform simulated RIA tests with prototypic test conditions is currently being evaluated by ANL personnel. TREAT was designed to accommodate test loops and vehicles installed for in-pile transient testing. During 40 years of TREAT operation and fuel testing and evaluation, experimenters have been able to demonstrate and determine the transient behavior of several types of fuel under a variety of test conditions. This experience led to an evolution of test methodology and techniques which can be employed to assess RIA behavior of LWR fuel. A pressurized water loop that will accommodate RIA testing of LWR and CANDU-type fuel has completed conceptual design. Preliminary calculations of transient characteristics and energy deposition into test rods during hypothetical TREAT RIA tests indicate that with the installation of a pressurized water loop, the facility is quite capable of performing prototypic RIA testing. Typical test scenarios indicate that a simulated RIA with a 72 msec FWHM pulse width and energy deposition of 1200 kJ/kg (290 cal/gm) is possible. Further control system enhancements would expand the capability to pulse widths as narrow as 40 msec. (author)

  5. The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

    1992-01-01

    A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

  6. Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1989-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

  7. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  8. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges - 15066

    International Nuclear Information System (INIS)

    Sabharwall, P.; O'Brien, J.E.; Yoon, S.J.; Sun, X.

    2015-01-01

    A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic, materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The 3 loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuits heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integrated System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 C. degrees), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF 4 ) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼ 450 C. degrees. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be

  9. Alteration in reactor installations (Unit 1 and 2 reactor facilities) in the Hamaoka Nuclear Power Station of The Chubu Electric Power Co., Inc. (report)

    International Nuclear Information System (INIS)

    1982-01-01

    A report by the Nuclear Safety Commission to the Ministry of International Trade and Industry concerning the alteration in Unit 1 and 2 reactor facilities in the Hamaoka Nuclear Power Station, Chubu Electric Power Co., Inc., was presented. The technical capabilities for the alteration of reactor facilities in Chubu Electric Power Co., Inc., were confirmed to be adequate. The safety of the reactor facilities after the alteration was confirmed to be adequate. The items of examination made for the confirmation of the safety are as follows: reactor core design (nuclear design, mechanical design, mixed reactor core), the analysis of abnormal transients in operation, the analysis of various accidents, the analysis of credible accidents for site evaluation. (Mori, K.)

  10. Annual report for RCRA groundwater monitoring projects at Hanford Site facilities for 1995

    International Nuclear Information System (INIS)

    Hartman, M.J.

    1996-02-01

    This report presents the annual hydrogeologic evaluation of 19 Resource Conservation and Recovery Act of 1976 facilities and 1 nonhazardous waste facility at the US Department of Energy's Hanford Site. Although most of the facilities no longer receive dangerous waste, a few facilities continue to receive dangerous waste constituents for treatment, storage, or disposal. The 19 Resource Conservation and Recovery Act facilities comprise 29 waste management units. Nine of the units are monitored under groundwater quality assessment status because of elevated levels of contamination indicator parameters. The impact of those units on groundwater quality, if any, is being investigated. If dangerous waste or waste constituents have entered groundwater, their concentration profiles, rate, and extent of migration are evaluated. Groundwater is monitored at the other 20 units to detect leakage, should it occur. This report provides an interpretation of groundwater data collected at the waste management units between October 1994 and September 1995. Groundwater quality is described for the entire Hanford Site. Widespread contaminants include nitrate, chromium, carbon tetrachloride, tritium, and other radionuclides

  11. Annual report for RCRA groundwater monitoring projects at Hanford site facilities for 1994

    International Nuclear Information System (INIS)

    1995-02-01

    This report presents the annual hydrogeologic evaluation of 19 Resource Conservation and Recovery Act of 1976 facilities and 1 nonhazardous waste facility at the U.S. Department of Energy's Hanford Site. Although most of the facilities no longer receive dangerous waste, a few facilities continue to receive dangerous waste constituents for treatment, storage, or disposal. The 19 Resource Conservation and Recovery Act facilities comprise 29 waste management units. Nine of the units are monitored under groundwater quality assessment status because of elevated levels of contamination indicator parameters. The impact of those units on groundwater quality, if any, is being investigated. If dangerous waste or waste constituents have entered groundwater, their concentration profiles, rate, and extent of migration are evaluated. Groundwater is monitored at the other 20 units to detect leakage, should it occur. This report provides an interpretation of groundwater data collected at the waste management units between October 1993 and September 1994. Groundwater quality is described for the entire Hanford Site. Widespread contaminants include nitrate, chromium, carbon tetrachloride, tritium, and other radionuclides

  12. Experimental facilities for PEC reactor design central channel test loop: CPC-1 - thermal shocks loop: CEDI

    International Nuclear Information System (INIS)

    Calvaresi, C.; Moreschi, L.F.

    1983-01-01

    PEC (Prova Elementi di Combustibile: Fuel Elements Test) is an experimental fast sodium-cooled reactor with a power of 120 MWt. This reactor aims at studying the behaviour of fuel elements under thermal and neutron conditions comparable with those existing in fast power nuclear facilities. Given the particular structure of the core, the complex operations to be performed in the transfer cell and the strict operating conditions of the central channel, two experimental facilities, CPC-1 and CEDI, have been designed as a support to the construction of the reactor. CPC-1 is a 1:1 scale model of the channel, transfer-cell and loop unit of the channel, whereas CEDI is a sodium-cooled loop which enables to carry out tests of isothermal endurance and thermal shocks on the group of seven forced elements, by simulating the thermo-hydraulic and mechanical conditions existing in the reactor. In this paper some experimental test are briefy discussed and some facilities are listed, both for the CPC-1 and for the CEDI. (Auth.)

  13. A midsize reactor facility - A regional resource for research and education

    International Nuclear Information System (INIS)

    Vernetson, W.G.

    1991-01-01

    The mission of the University of Florida Training Reactor (UFTR) is to serve the regional needs of Florida and the Southeast for access to quality reactor usage. Well-advertised capabilities of the facility support diversified usages that include education, training, research, service, and public information programs to address the needs of a broad spectrum of users ranging from high school students and teachers, to university researchers, and even the occasional service user. Despite the midsize power of the facility, the UFTR's status as the only nonpower reactor within 350 miles in one of our largest states means that it is uniquely situated to contribute in these various areas in ways usually reserved for larger facilities. Nine state universities and a well-developed community college system in addition to private schools and a growing complement of progressive high schools assure a broad-based user community. The key to accomplishing mission objectives is to continue diversification and improvement of both the reactor and associated experimental capabilities to meet the needs of this user community

  14. TIT reactor laboratory course using JAERI and PNC large experimental facilities

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Obara, Toru; Ohtani, Nobuo.

    1995-01-01

    This report is presented on a reactor laboratory course for graduate students using large facilities in national laboratories in Japan. A reactor laboratory course is offered every summer since 1990 for all graduate students in the Nuclear Engineering Course in Tokyo Institute of Technology (TIT), where the students can choose one of the experiments prepared at Japan Atomic Energy Research Institute (JAERI), Power Reactor and Nuclear Fuel Development Corporation (PNC) and Research Reactor Institute, Kyoto University (KUR). Both JAERI and PNC belong to Science and Technology Agency (STA). This is the first university curriculum of nuclear engineering using the facilities owned by the STA laboratories. This type of collaboration is promoted in the new Long-Term Program for Research, Development and Utilization of Nuclear Energy adopted by Atomic Energy Commission. Most students taking this course reported that they could learn so much about reactor physics and engineering in this course and the experiment done in large laboratory was a very good experience for them. (author)

  15. Shadow corrosion testing in the INCA facility in the Studsvik R2 reactor

    International Nuclear Information System (INIS)

    Nystrand, A.C.; Lassing, A.

    1999-01-01

    Shadow corrosion is a phenomenon which occurs when zirconium alloys are in contact with or in proximity to other metallic objects in a boiling water reactor environment (BWR, RBMK, SGHWR etc.). An enhanced corrosion occurs on the zirconium alloy with the appearance of a 'shadow' of the metallic object. The magnitude of the shadow corrosion can be significant, and is potentially limiting for the lifetime of certain zirconium alloy components in BWRs and other reactors with a similar water chemistry. In order to evaluate the suitability of the In-Core Autoclave (INCA) in the Studsvik R2 materials testing reactor as an experimental facility for studying shadow corrosion, a demonstration test has been performed. A number of test specimens consisting of Zircaloy-2 tubing in contact with Inconel were exposed in an oxidising water chemistry. Some of the specimens were placed within the reactor core and some above the core. The conclusion of this experiment after post irradiation examination is that it is possible to use the INCA facility in the Studsvik R2 reactor to develop a significant level of shadow corrosion after only 800 hours of irradiation. (author)

  16. Joint Assessment of ETRR-2 Research Reactor Operations Program, Capabilities, and Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Bissani, M; O' Kelly, D S

    2006-05-08

    A joint assessment meeting was conducted at the Egyptian Atomic Energy Agency (EAEA) followed by a tour of Egyptian Second Research Reactor (ETRR-2) on March 22 and 23, 2006. The purpose of the visit was to evaluate the capabilities of the new research reactor and its operations under Action Sheet 4 between the U.S. DOE and the EAEA, ''Research Reactor Operation'', and Action Sheet 6, ''Technical assistance in The Production of Radioisotopes''. Preliminary Recommendations of the joint assessment are as follows: (1) ETRR-2 utilization should be increased by encouraging frequent and sustained operations. This can be accomplished in part by (a) Improving the supply-chain management for fresh reactor fuel and alleviating the perception that the existing fuel inventory should be conserved due to unreliable fuel supply; and (b) Promulgating a policy for sample irradiation priority that encourages the use of the reactor and does not leave the decision of when to operate entirely at the discretion of reactor operations staff. (2) Each experimental facility in operation or built for a single purpose should be reevaluated to focus on those that most meet the goals of the EAEA strategic business plan. Temporary or long-term elimination of some experimental programs might be necessary to provide more focused utilization. There may be instances of emerging reactor applications for which no experimental facility is yet designed or envisioned. In some cases, an experimental facility may have a more beneficial use than the purpose for which it was originally designed. For example, (a) An effective Boron Neutron Capture Therapy (BNCT) program requires nearby high quality medical facilities. These facilities are not available and are unlikely to be constructed near the Inshas site. Further, the BNCT facility is not correctly designed for advanced research and therapy programs using epithermal neutrons. (b) The ETRR-2 is frequently operated to

  17. Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

    2007-11-01

    This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

  18. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® Reactor

    International Nuclear Information System (INIS)

    Schickler, R.A.; Marcum, W.R.; Reese, S.R.

    2013-01-01

    Highlights: • The Oregon State TRIGA ® Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA ® Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least-squares technique. The quantification of

  19. Modernization of the facilities of the TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Mendez T, D.; Flores C, J.

    2016-09-01

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

  20. The LEU target development and conversion program for the MAPLE reactors and new processing facility

    International Nuclear Information System (INIS)

    Malkoske, G.R.

    2003-01-01

    The availability of isotope grade, Highly Enriched Uranium (HEU), from the United States for use in the manufacture of targets for molybdenum-99 production in AECL's NRU research reactor has been a key factor to enable MDS Nordion to develop a reliable, secure supply of medical isotopes for the international nuclear medicine community. The molybdenum extraction process from HEU targets is a proven and established method that has reliably produced medical isotopes for several decades. The HEU process provides predictable, consistent yields for our high-volume, molybdenum-99 production. Other medical isotopes such as I-131 and Xe-133, which play an important role in nuclear medicine applications, are also produced from irradiated HEU targets as a by-product of the molybdenum-99 process. To ensure a continued reliable and timely supply of medical isotopes, MDS Nordion is completing the commissioning of two MAPLE reactors and an associated isotope processing facility (the New Processing Facility). The new MAPLE facilities, which will be dedicated exclusively to medical isotope production, will provide an essential contribution to a secure, robust global healthcare system. Design and construction of these facilities has been based on a life cycle management philosophy for the isotope production process. This includes target irradiation, isotope extraction and waste management. The MAPLE reactors will operate with Low Enriched Uranium (LEU) fuel, a significant contribution to the objectives of the RERTR program. The design of the isotope production process in the MAPLE facilities is based on an established process - extraction of isotopes from HEU target material. This is a proven technology that has been demonstrated over more than three decades of operation. However, in support of the RERTR program and in compliance with U.S. legislation, MDS Nordion has undertaken a LEU Target Development and Conversion Program for the MAPLE facilities. This paper will provide an