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Sample records for davis besse-1 reactor

  1. Davis-Besse uncertainty study

    International Nuclear Information System (INIS)

    Davis, C.B.

    1987-08-01

    The uncertainties of calculations of loss-of-feedwater transients at Davis-Besse Unit 1 were determined to address concerns of the US Nuclear Regulatory Commission relative to the effectiveness of feed and bleed cooling. Davis-Besse Unit 1 is a pressurized water reactor of the raised-loop Babcock and Wilcox design. A detailed, quality-assured RELAP5/MOD2 model of Davis-Besse was developed at the Idaho National Engineering Laboratory. The model was used to perform an analysis of the loss-of-feedwater transient that occurred at Davis-Besse on June 9, 1985. A loss-of-feedwater transient followed by feed and bleed cooling was also calculated. The evaluation of uncertainty was based on the comparisons of calculations and data, comparisons of different calculations of the same transient, sensitivity calculations, and the propagation of the estimated uncertainty in initial and boundary conditions to the final calculated results

  2. Technical evaluation of the alternate to the keylock control to the bypass valves for the Davis-Besse nuclear power plant, Unit 1

    International Nuclear Information System (INIS)

    Ibarra, J.G.

    1979-09-01

    This report documents the technical evaluation of the alternate to the keylock control to the bypass valves for the Davis-Besse nuclear power plant, Unit 1. The review criteria are inferred from the NRC Reactor Safety Study (WASH-1400) and the Safety Evaluation Report for Davis-Besse. This report is supplied as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program being conducted for the US Nuclear Regulatory Commission by Lawrence Livermore Laboratory

  3. Life extension program initiation at Davis-Besse nuclear power station

    International Nuclear Information System (INIS)

    Staudinger, Deborah K.

    1991-01-01

    Davis-Besse is a 900 MW Babcock and Wilcox designed plant located in Northwest Ohio. Effective December 31, 1990, the construction period was recovered making the current license expiration 2007. The economic effects of this extension reduced the depreciation expense for 1990 by $9,790,000 and increased earnings per share by $.04. This positive impact has resulted in an evaluation of pursuing license renewal for Davis-Besse in accordance with the proposed rule on license renewal (10CFR54 'Requirements for renewal of operating licenses for nuclear power plants'). This paper reviews preliminary efforts to evaluate these actions and summarizes strategies planned to ensure continued operation of Davis-Besse remains a viable option for base load generation for Toledo Edison. (author)

  4. Loss of main and auxiliary feedwater event at the Davis-Besse Plant on June 9, 1985

    International Nuclear Information System (INIS)

    1985-07-01

    On June 9, 1985, Toledo Edison Company's Davis-Besse Nuclear Power Plant, located in Ottawa County, Ohio, experienced a partial loss of feedwater while the plant was operating at 90% power. Following a reactor trip, a loss of all feedwater occurred. The event involved a number of equipment malfunctions and extensive operator actions, including operator actions outside the control room. Several operator errors also occurred during the event. This report documents the findings of an NRC Team sent to Davis-Besse by the NRC Executive Director for Operations in conformance with the staff-proposed Incident Investigation Program

  5. Report of the independent Ad Hoc Group for the Davis-Besse incident

    International Nuclear Information System (INIS)

    1986-06-01

    The Nuclear Regulatory Commission established an independent Ad Hoc Group in January 1986 to review issues subsequent to a complete loss of feedwater event at Davis-Besse Nuclear Power Station on June 9, 1985, including the NRC Incident Investigation Team (IIT) investigation of that event. The Commission asked the Group to identify additional lessons that might be learned and from these to make recommendations to improve NRC oversight of reactor licensees. To fulfill its charter, the Ad Hoc Group examined the following: (1) pre-event interactions between the licensee and NRC concerning reliability of the auxiliary feedwater system and associated systems; (2) pre-event probabilistic assessments of the reliability of plant safety systems, NRC's review of them, and their use in regulatory decisionmaking; (3) licensee management, operation and maintenance programs as they may have contributed to equipment failures and NRC oversight of such programs; and (4) the mandate, capabilities of members, operation, and results of the NRC Davis-Besse IIT, and the use to which its report was put by the regulatory staff

  6. Safety Evaluation Report related to the restart of Davis-Besse Nuclear Power Station, Unit 1, following the event of June 9, 1985 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    1986-06-01

    On June 9, 1985, the Davis-Besse Nuclear Power Station, operated by the Toledo Edison Company, experienced a partial loss of main feedwater while the plant was at 90% power. The ensuing reactor trip was followed by spurious isolation of the steam geneators which initiated a chain of events involving a number of equipment malfunctions and several operator errors ultimately interrupting all feedwater for a short period of time. By the time operators were able to restore feedwater, both steam generators had dried out. A letter from the Director of the Office of Nuclear Reactor Regulation, pursuant to 10 CFR 50.54(f) of the Commission's regulations, confirmed that the Davis-Besse facility would not be restarted without NRC approval. The letter also requested that Toledo Edison submit its program for resolving numerous concerns identified by the staff. In response, the license submitted the Davis-Besse Course of Action report. The staff has reviewed that document and other supporting material submitted by the licensee; the staff's evaluation of that information is presented in this report

  7. Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

    International Nuclear Information System (INIS)

    Wesley, D.A.

    1991-01-01

    A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

  8. Developing excellence awareness at Davis-Besse through engineering quality management

    International Nuclear Information System (INIS)

    Gaudette, M.R.; Lash, J.H.; Haiman, D.L.

    1989-01-01

    At Davis-Besse nuclear power station engineering quality management includes a variety of chartered functions whose ultimate objective is to improve product and service quality and process efficiency. These functions were assigned in late 1988 to engineering assurance personnel, a section of the engineering department which reports directly to the engineering director but is independent of design activities. This independence ensures objectivity and allows the improvement process to span functional areas so that changes made in one engineering section do not negatively impact the activities of another section. The engineering quality management functions performed by the engineering assurance group are summarized. Engineering quality management at Davis-Besse has increased the degree of excellence evident in engineering products and services

  9. Root-cause analysis - An essential culture at Davis-Besse

    International Nuclear Information System (INIS)

    Garver, R.G.; Gerren, D.W.; Jain, S.C.

    1990-01-01

    An ingrained cultural attitude toward diligent pursuit of the root causes of plant anomalies and equipment malfunctions, and effective implementation of corrective actions is essential to achieve and maintain desired safety and performance standards at a nuclear power plant. At the Davis-Besse nuclear power station, a demonstrated management commitment to these actions, coupled with an effective root-cause training program has made root-cause analysis an important engineering function. The dedication of engineering and maintenance department resources to problem investigation and troubleshooting, root-cause analysis, and corrective action implementation has eliminated several complex operational problems. Reactor trips, unplanned challenges to safety systems, and unplanned plant transients have been significantly reduced as a result. The benefits of these plant performance improvements far outweigh the expense of these resources

  10. Technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1

    International Nuclear Information System (INIS)

    Selan, J.C.

    1981-07-01

    This report documents the technical evaluation of the noise and isolation testing of the safety features actuation system at the Davis Besse Nuclear Power Station, Unit 1. The tests were to verify that faults on the non-Class 1E circuits would not propagate to the Class 1E circuits and degrade them below acceptable levels. The tests conducted demonstrated that the safety features actuation system did not degrade below acceptable levels nor was the system's ability to perform its protective functions affected

  11. Pressure-dependent fragilities for piping components: Pilot study on Davis-Besse Nuclear Power Station

    International Nuclear Information System (INIS)

    Wesley, D.A.; Nakaki, D.K.; Hadidi-Tamjed, H.; Kipp, T.R.

    1990-10-01

    The capacities of four, low-pressure fluid systems to withstand pressures and temperatures above the design levels were established for the Davis-Besse Nuclear Power Station. The results will be used in evaluating the probability of plant damage from Interfacing System Loss of Coolant Accidents (ISLOCA) as part of the probabilistic risk assessment of the Davis-Besse nuclear power station undertaken by EG ampersand G Idaho, Inc. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The probabilities of failure, as a function of internal pressure, are evaluated as well as the variabilities associated with them. Leak rates or leak areas are estimated for the controlling modes of failure. The pressure capacities for the pipes and vessels are evaluated using limit-state analyses for the various failure modes considered. The capacities are dependent on several factors, including the material properties, modeling assumptions, and the postulated failure criteria. The failure modes for gasketed-flange connections, valves, and pumps do not lend themselves to evaluation by conventional structural mechanics techniques and evaluation must rely primarily on the results from ongoing gasket research test programs and available vendor information and test data. 21 refs., 7 figs., 52 tabs

  12. An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

    International Nuclear Information System (INIS)

    Di Salvo, R.; Leonard, M.T.; Wreathall, J.

    1986-01-01

    An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

  13. Aerial radiological survey of the Davis-Besse Nuclear power Station and surrounding area, Oak Harbor, Ohio. Date of survey: 26-29 May 1980

    International Nuclear Information System (INIS)

    Hilton, L.K.

    1980-12-01

    An airborne radiological survey of a 130 km 2 area centered on the Davis-Besse Nuclear Power Station was made 26-29 May 1980. Count rates observed at 90 m altitude were converted to exposure rates at 1 m above the ground and are presented in the form of an isopleth map. Detected radioisotopes and their associated gamma ray exposure rates were consistent with that expected from normal background emitters, except directly over the station

  14. Ranking of risk significant components for the Davis-Besse Component Cooling Water System

    International Nuclear Information System (INIS)

    Seniuk, P.J.

    1994-01-01

    Utilities that run nuclear power plants are responsible for testing pumps and valves, as specified by the American Society of Mechanical Engineers (ASME) that are required for safe shutdown, mitigating the consequences of an accident, and maintaining the plant in a safe condition. These inservice components are tested according to ASME Codes, either the earlier requirements of the ASME Boiler and Pressure Vessel Code, Section XI, or the more recent requirements of the ASME Operation and Maintenance Code, Section IST. These codes dictate test techniques and frequencies regardless of the component failure rate or significance of failure consequences. A probabilistic risk assessment or probabilistic safety assessment may be used to evaluate the component importance for inservice test (IST) risk ranking, which is a combination of failure rate and failure consequences. Resources for component testing during the normal quarterly verification test or postmaintenance test are expensive. Normal quarterly testing may cause component unavailability. Outage testing may increase outage cost with no real benefit. This paper identifies the importance ranking of risk significant components in the Davis-Besse component cooling water system. Identifying the ranking of these risk significant IST components adds technical insight for developing the appropriate test technique and test frequency

  15. Final environmental statement: Related to the operation of Davis-Besse Nuclear Power Station, Unit 1 (Docket No. 50-346)

    International Nuclear Information System (INIS)

    1975-10-01

    The proposed action is the issuance of an operating license to the Toledo Edison Company and the Cleveland Electric Illuminating Company for the startup and operation of the Davis-Besse Nuclear Power Station Unit 1 (the station) located near Port Clinton in Ottawa County, Ohio. The total site area is 954 acres of which 160 acres have been removed from production of grain crops and converted to industrial use. Approximately 600 acres of the area is marshland which will be maintained as a wildlife refuge. The disturbance of the lake shore and lake bottom during construction of the station water intake and discharge pipes resulted in temporary turbidity, silting, and destruction of bottom organisms. Since completion of these activities, evidence of improvement in turbidity and transparency measurements, and the reestablishment of the bottom organism has been obtained. The cooling tower blowdown and service water which the station discharges to Lake Erie, via a submerged jet, will be heated no more than 20/degrees/F above the ambient lake water temperature. Although some small fish and plankton in the discharge water plume will be disabled as a result of thermal shock, exposure to chlorine and buffeting, few adult fish will be affected. The thermal plume resulting from the maximum thermal discharge is calculated to have an area of less than one acre within the 3/degrees/F isotherm (above lake ambient). Approximately 101 miles of transmission lines have been constructed, primarily over existing farmland, requiring about 1800 acres of land for the rights-of-way. Land use will essentially be unchanged since only the land required for the base of the towers is removed from production. Herbicides will not be used to maintain the rights-of-way. 14 figs., 34 refs

  16. Development of a reactor coolant pump monitoring and diagnostic system. Progress report, June 1982-July 1983

    International Nuclear Information System (INIS)

    Morris, D.J.; Sommerfield, G.A.

    1983-12-01

    The quality of operating data has been insufficient to allow proper evaluation of theoretical reactor coolant (RC) pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables: The rotordynamic behavior of the pump shaft and related components, the internal conditions and performance of the seals, and the plant or pump operating environment (controlled by the plant operator). Interrelationships between these areas will be developed during the data collection task, scheduled to begin in October 1983 (for a full fuel cycle at Davis-Besse). This report describes system software and hardware development, testing, and installation work performed during this period. Also described is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  17. Prototype vibration measurement program for reactor internals (177-fuel assembly plant). Supplement 1

    International Nuclear Information System (INIS)

    Simonis, J.C.; Post, R.C.; Thoren, D.E.

    1976-08-01

    The surveillance specimen holder tubes installed in the Babcock and Wilcox 177-fuel assembly plants have been redesigned. The structural adequacy of this design has been verified through extensive analysis. The design adequacy will be further confirmed by measuring the vibrational response of the surveillance specimen holder tube during normal and transient flow operation. This report describes the vibration measurement program that will be conducted at Toledo Edison's Davis Besse 1 site

  18. Status of University of Cincinnati reactor-site nuclear engineering graduate programs

    International Nuclear Information System (INIS)

    Anno, J.N.; Christenson, J.M.; Eckart, L.E.

    1993-01-01

    The University of Cincinnati (UC) nuclear engineering program faculty has now had 12 yr of experience in delivering reactor-site educational programs to nuclear power plant technical personnel. Currently, with the sponsorship of the Toledo-Edison Company (TED), we are conducting a multiyear on-site graduate program with more than 30 participants at the Davis-Besse nuclear power plant. The program enables TED employees with the proper academic background to earn a master of science (MS) degree in nuclear engineering (mechanical engineering option). This paper presents a brief history of tile evolution of UC reactor-site educational programs together with a description of the progress of the current program

  19. Antimatter and Dark Matter Search in Space: BESS-Polar Results

    Science.gov (United States)

    Mitchell, John W.; Yamamoto, Akira

    2009-01-01

    The apex of the Balloon-borne Experiment with a Superconducting Spectrometer program was reached with the Antarctic flight of BESS-Polar II, during the 2007-2008 Austral Summer, that obtained 24.5 days of data on over 4.7 billion cosmic-ray events. The US-Japan BESS Collaboration uses elementary particle measurements to study the early Universe and provides fundamental data on the spectra of light cosmic-ray elements and isotopes. BESS measures the energy spectra of cosmic-ray antiprotons to investigate signatures of possible exotic sources, such as dark-matter candidates, and searches for heavier anti-nuclei that might reach Earth from antimatter domains formed during symmetry breaking processes in the early Universe. Since 1993, BESS has carried out eleven high-latitude balloon flights, two of long duration, that together have defined the study of antiprotons below about 4 GeV, provided standard references for light element and isotope spectra, and set the most sensitive limits on the existence of anti-deuterons and anti-helium, The BESS-Polar II flight took place at Solar Minimum, when the sensitivity of the low-energy antiproton measurements to a primary source is greatest. The rich BESS-Polar II dataset more than doubles the combined data from all earlier BESS flights and has 10-20 times the statistics of BESS data from the previous Solar Minimum. Here, we summarize the scientific results of BESS program, focusing on the results obtained using data from the long-duration flights of BESS-Polar I (2004) and BESS-Polar II.

  20. Sway Area and Velocity Correlated With MobileMat Balance Error Scoring System (BESS) Scores.

    Science.gov (United States)

    Caccese, Jaclyn B; Buckley, Thomas A; Kaminski, Thomas W

    2016-08-01

    The Balance Error Scoring System (BESS) is often used for sport-related concussion balance assessment. However, moderate intratester and intertester reliability may cause low initial sensitivity, suggesting that a more objective balance assessment method is needed. The MobileMat BESS was designed for objective BESS scoring, but the outcome measures must be validated with reliable balance measures. Thus, the purpose of this investigation was to compare MobileMat BESS scores to linear and nonlinear measures of balance. Eighty-eight healthy collegiate student-athletes (age: 20.0 ± 1.4 y, height: 177.7 ± 10.7 cm, mass: 74.8 ± 13.7 kg) completed the MobileMat BESS. MobileMat BESS scores were compared with 95% area, sway velocity, approximate entropy, and sample entropy. MobileMat BESS scores were significantly correlated with 95% area for single-leg (r = .332) and tandem firm (r = .474), and double-leg foam (r = .660); and with sway velocity for single-leg (r = .406) and tandem firm (r = .601), and double-leg (r = .575) and single-leg foam (r = .434). MobileMat BESS scores were not correlated with approximate or sample entropy. MobileMat BESS scores were low to moderately correlated with linear measures, suggesting the ability to identify changes in the center of mass-center of pressure relationship, but not higher-order processing associated with nonlinear measures. These results suggest that the MobileMat BESS may be a clinically-useful tool that provides objective linear balance measures.

  1. The BESS Search for Cosmic-Ray Antiproton Origins and for Cosmological Antimatter

    Science.gov (United States)

    Mitchell, John; Yamamoto, Akira

    2009-01-01

    The apex of the Balloon-borne Experiment with a Superconducting Spectrometer (BESS) program was reached with the Antarctic flight of BESS-Polar II, during the 2007-2008 Austral Summer, that obtained 24.5 days of data on over 4.7 billion cosmic-ray events. The US-Japan BESS Collaboration uses elementary particle measurements to study the early Universe and provides fundamental data on the spectra of light cosmic-ray elements and isotopes. BESS measures the energy spectra of cosmic-ray antiprotons to investigate signatures of possible exotic sources, such as dark-matter candidates, and searches for heavier antinuclei that might reach Earth from antimatter domains formed during symmetry breaking processes in the early Universe. Since 1993, BESS has carried out eleven high-latitude balloon flights, two of long duration, that together have defined the study of antiprotons below about 4 GeV, provided standard references for light element and isotope spectra, and set the most sensitive limits on the existence of antideuterons and antihelium. The BESS-Polar II flight took place at Solar Minimum, when the sensitivity of the low-energy antiproton measurements to a primary source is greatest. The rich BESS-Polar II dataset more than doubles the combined data from all earlier BESS flights and has 10-20 times the statistics of BESS data from the previous Solar Minimum. Here, we summarize the scientific results of BESS program, focusing on the results obtained using data from the long-duration flights of BESS-Polar I (2004) and BESS-Polar II.

  2. Georges Besse 2. A new era for enrichment

    International Nuclear Information System (INIS)

    2009-01-01

    Since 1978, AREVA group subsidiary EURODIF's Georges Besse plant has been using gaseous diffusion to enrich uranium and meet the requirements of electricity utilities. Georges Besse II plant, which will use far less electricity, will replace George Besse not so long. The Georges Besse II plant is located on the Tricastin nuclear site in southern France. AREVA ensure delivery of uranium enrichment services in accordance with customer expectations. AREVA obtained the right to use URENCO's centrifugation technology on July 3, 2006. This is the process to be used at Georges Besse II. The new uranium enrichment plant will comprise two enrichment units, with a total production capacity of 7.5 million separative work units, which can be extended to 11 million if needed. Each enrichment unit will include: a Centrifuge Assembly Building (CAB), a Centrifuge Utility Building (CUB) with offices and control room, annexes for purification, supply and extraction of uranium hexafluoride (UF 6 ), modules containing the halls that house the centrifuge cascades. The modular design of the Georges Besse II plant will allow production in the first cascade where as the others cascades are build. The first cascade should be operational in the first half of 2009. At an overall cost of 3 billion euros, this project is one of the largest investments of the decade in France On November 24, 2003, AREVA and URENCO signed an agreement under which AREVA would buy 50% of the shares in the Enrichment Technology Company (ETC), which designs and manufactures centrifuges

  3. Break location influence in pressure vessel SBLOCA scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Querol, Andrea; Gallardo, Sergio; Verdú, Gumersindo, E-mail: anquevi@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), Universitat Politècnica de València (Spain)

    2017-07-01

    The inspections performed in Davis Besse and in the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around of two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused in simulating different hypothetical accidental scenarios in the PV using the thermalhydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios. (author)

  4. Search for Cosmic-Ray Antiproton Origins and for Cosmological Antimatter with BESS

    Science.gov (United States)

    Yamamoto, A.; Mitchell, J. W.; Yoshimura, K.; Abe, K.; Fuke, H.; Haino, S.; Hams, T.; Hasegawa, M.; Horikoshi, A.; Itazaki, A.; hide

    2011-01-01

    The balloon-borne experiment with a superconducting spectrometer (BESS) has performed cosmic-ray observations as a US-Japan cooperative space science program, and has provided fundamental data on cosmic rays to study elementary particle phenomena in the early Universe. The BESS experiment has measured the energy spectra of cosmic-ray antiprotons to investigate signatures of possible exotic origins such as dark matter candidates or primordial black holes. and searched for heavier antinuclei that might reach Earth from antimatter domains formed in the early Universe. The apex of the BESS program was reached with the Antarctic flight of BESS-Polar II, during the 2007- 2008 Austral Summer, that obtained over 4.7 billion cosmic-ray events from 24.5 days of observation. The flight took place at the expected solar minimum, when the sensitivity of the low-energy antiproton measurements to a primary source is greatest. Here, we report the scientific restults, focusing on the long-duration flights of BESS-Polar I (2004) and BESS-Polar II (2007-2008).

  5. The Bess Investigation of the Origin of Cosmic-ray Antiprotons and Search for Cosmological Antimatter

    Science.gov (United States)

    Mitchell, John; Yamamoto, Akira; Yoshimura, Koji; Makida, Yasuhiro; Matsuda, Shinya; Hasegawa, Masaya; Horikoshi, Atsushi; Tanaka,Ken-ichi; Suzuki, Junichi; Nishimura, Jun; hide

    2008-01-01

    The Balloon-borne Experiment with a Superconducting Spectrometer (BESS) collaboration has made precise measurements of the spectra of cosmic ray antiprotons and light nuclei and conducted a sensitive search for antinuclei. Ten BESS high-latitude flights, eight from Canada and two from Antarctica, span more than a Solar cycle between 1993 and 2007/2008. BESS measurements of low-energy antiprotons constrain candidate models for dark matter including the possible signature of primordial black hole evaporation. The stringent BESS measurements of antiprotons and the elemental and isotopic spectra of H and He provide strong constraints on models of cosmic-ray transport in the Galaxy and Solar System. BESS has also reported the first antideuterium upper limit. BESS employs a superconducting magnetic-rigity spectrometer with time-of-flight and aerogel Cherenkov detectors to identify incident particles by charge, charge sign, mass, and energy. The BESS-Polar long-duration instrument has reduced lower energy limit of 100 MeV (top of the atmosphere) to increase its sensitivity to possible primary antiproton sources. BESS-Polar II was rebuilt with extended magnet lifetime, improved detector and electronic performance, and greater data storage capacity. It was flown fro Antarctica December 2007-January 2008, recording about 4.6 bission events during 24.5 days at float altitude with the magnet on. During the flight the influence of a high-speed stream in the Solar wind was observed. Details of the BESS-Polar II instrument and flight performance are reported elsewhere at this conference. The successful BESS-Polar II flight at Solar minimum is especially important. Most cosmic-ray antiprotons are secondary products of nuclear interactions of primary cosmic-ray nuclei with the interstellar gas, giving a spectrum that peaks at about 2 GeV and falls rapidly to higher and lower energies. However, BESS data taken in the previous Solar minimum show a small excess over secondary

  6. Development of a reactor-coolant-pump monitoring and diagnostic system. Semi-annual progress report, December 1981-May 1982

    International Nuclear Information System (INIS)

    Morris, D.J.; Gabler, H.C.

    1982-10-01

    Reactor coolant (RC) pump seal failures have resulted in excessive leakage of primary coolant into reactor containment buildings. In some cases, high levels of airborne activity and surface contamination following these failures have necessitated extensive cleanup efforts and personnel radiation exposure. Unpredictable pump seal performance has also caused forced outages and frequent maintenance. The quality of operating data has been insufficient to allow proper evaluation of theoretical RC pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables. This report describes system software and hardware development, testing, and installation work performed during the period of December 1981 through May 1982. Also described herein is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  7. SOLAR MODULATION OF THE LOCAL INTERSTELLAR SPECTRUM WITH VOYAGER 1 , AMS-02, PAMELA , AND BESS

    Energy Technology Data Exchange (ETDEWEB)

    Corti, C.; Bindi, V.; Consolandi, C.; Whitman, K., E-mail: corti@hawaii.edu [Physics and Astronomy Department, University of Hawaii at Manoa, Honolulu, HI 96822 (United States)

    2016-09-20

    In recent years, the increasing precision of direct cosmic rays measurements opened the door to high-sensitivity indirect searches of dark matter and to more accurate predictions for radiation doses received by astronauts and electronics in space. The key ingredients in the study of these phenomena are the knowledge of the local interstellar spectrum (LIS) of galactic cosmic rays and the understanding of how the solar modulation affects the LIS inside the heliosphere. Voyager 1 , AMS-02, PAMELA , and BESS measurements of proton and helium fluxes provide valuable information, allowing us to shed light on the shape of the LIS and the details of the solar modulation during solar cycles 22-24. A new parametrization of the LIS is presented, based on the latest data from Voyager 1 and AMS-02. Using the framework of the force-field approximation, the solar modulation parameter is extracted from the time-dependent fluxes measured by PAMELA and BESS . A modified version of the force-field approximation with a rigidity-dependent modulation parameter is introduced, yielding better fits than the force-field approximation. The results are compared with the modulation parameter inferred by neutron monitors.

  8. THE Be STAR SPECTRA (BeSS) DATABASE

    International Nuclear Information System (INIS)

    Neiner, C.; De Batz, B.; Cochard, F.; Floquet, M.; Mekkas, A.; Desnoux, V.

    2011-01-01

    Be stars vary on many timescales, from hours to decades. A long time base of observations to analyze certain phenomena in these stars is therefore necessary. Collecting all existing and future Be star spectra into one database has thus emerged as an important tool for the Be star community. Moreover, for statistical studies, it is useful to have centralized information on all known Be stars via an up-to-date catalog. These two goals are what the Be Star Spectra (BeSS, http://basebe.obspm.fr) database proposes to achieve. The database contains an as-complete-as-possible catalog of known Be stars with stellar parameters, as well as spectra of Be stars from all origins (any wavelength, any epoch, any resolution, etc.). It currently contains over 54,000 spectra of more than 600 different Be stars among the ∼2000 Be stars in the catalog. A user can access and query this database to retrieve information on Be stars or spectra. Registered members can also upload spectra to enrich the database. Spectra obtained by professional as well as amateur astronomers are individually validated in terms of format and science before being included in BeSS. In this paper, we present the database itself as well as examples of the use of BeSS data in terms of statistics and the study of individual stars.

  9. Trial application of guidelines for nuclear plant response to an earthquake

    International Nuclear Information System (INIS)

    Schmidt, W.; Oliver, R.; O'Connor, W.

    1993-09-01

    Guidelines have been developed to assist nuclear plant personnel in the preparation of earthquake response procedures for nuclear power plants. These guidelines are published in EPRI report NP-6695, ''Guidelines for Nuclear Plant Response to an Earthquake,'' dated December 1989. This report includes two sets of nuclear plant procedures which were prepared to implement the guidelines of EPRI report NP-6695. The first set were developed by the Toledo Edison Company Davis-Besse plant. Davis-Besse is a pressurized water reactor (PWR) and contains relatively standard seismic monitoring instrumentation typical of many domestic nuclear plants. The second set of procedures were prepared by Yankee Atomic Electric Company for the Vermont Yankee facility. This plant is a boiling water reactor (BWR) with state-of-the-art seismic monitoring and PC-based data processing equipment, software developed specifically to implement the OBE Exceedance Criterion presented in EPRI report NP-5930, ''A Criterion for Determining Exceedance of the operating Basis Earthquake.'' The two sets of procedures are intended to demonstrate how two different nuclear utilities have interpreted and applied the EPRI guidance given in report NP-6695

  10. Review of nuclear power reactor coolant system leakage events and leak detection requirements

    International Nuclear Information System (INIS)

    Chokshi, N.C.; Srinivasan, M.; Kupperman, D.S.; Krishnaswamy, P.

    2005-01-01

    In response to the vessel head event at the Davis-Besse reactor, the U.S. Nuclear Regulatory Commission (NRC) formed a Lessons Learned Task Force (LLTF). Four action plans were formulated to respond to the recommendations of the LLTF. The action plans involved efforts on barrier integrity, stress corrosion cracking (SCC), operating experience, and inspection and program management. One part of the action plan on barrier integrity was an assessment to identify potential safety benefits from changes in requirements pertaining to leakage in the reactor coolant system (RCS). In this effort, experiments and models were reviewed to identify correlations between crack size, crack-tip-opening displacement (CTOD), and leak rate in the RCS. Sensitivity studies using the Seepage Quantification of Upsets In Reactor Tubes (SQUIRT) code were carried out to correlate crack parameters, such as crack size, with leak rate for various types of crack configurations in RCS components. A database that identifies the leakage source, leakage rate, and resulting actions from RCS leaks discovered in U.S. light water reactors was developed. Humidity monitoring systems for detecting leakage and acoustic emission crack monitoring systems for the detection of crack initiation and growth before a leak occurs were also considered. New approaches to the detection of a leak in the reactor head region by monitoring boric-acid aerosols were also considered. (authors)

  11. A comparative evaluation of emerging methods for errors of commission based on applications to the Davis-Besse (1985) event

    International Nuclear Information System (INIS)

    Reer, B.; Dang, V.N.; Hirschberg, S.; Straeter, O.

    1999-12-01

    In considering the human role in accidents, the classical PSA methodology applied today focuses primarily on the omissions of actions required of the operators at specific points in the scenario models. A practical, proven methodology is not available for systematically identifying and analyzing the scenario contexts in which the operators might perform inappropriate actions that aggravate the scenario. As a result, typical PSA's do not comprehensively treat these actions, referred to as errors of commission (EOCs). This report presents the results of a joint project of the Paul Scherrer Institut (PSI, Villigen, Switzerland) and the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Garching, Germany) that examined some methods recently proposed for addressing the EOC issue. Five methods were investigated: 1 ) ATHEANA, 2) the Borssele screening methodology. 3) CREAM, 4) CAHR, and 5) CODA. In addition to a comparison of their scope, basic assumptions, and analytical approach, the methods were each applied in the analysis of PWR Loss of Feedwater scenarios based on the 1985 Davis-Besse event, in which the operator response included actions that can be categorized as EOCs. The aim was to compare how the methods consider a concrete scenario in which EOCs have in fact been observed. These case applications show how the methods are used in practical terms and constitute a common basis for comparing the methods and the insights that they provide. The identification of the potentially significant EOCs to be analysed in the PSA is currently the central problem for their treatment. The identification or search scheme has to consider an extensive set of potential actions that the operators may take. These actions may take place instead of required actions, for example, because the operators fail to assess the plant state correctly, or they may occur even when no action is required. As a result of this broad search space, most methodologies apply multiple schemes to

  12. A Comparison of Balance and Postural Stability Assessment Tools: BESS Versus NeuroCom Balance Manager

    OpenAIRE

    Joliffe, Jamie

    2012-01-01

    Postural stability assessment tools are one of the many ways concussions can be assessed and return to play decisions can be made; two of which are the Balance Error Scoring System (BESS) and force plate technology. OBJECTIVE: Validate the modified BESS used by Utah State University by comparing it to equivalent tests on the NeuroCom Balance Manager System. METHODS: 114 current or previous Utah State football players ranging in age from 18-24. Each athlete conducted a baseline BESS test durin...

  13. A comparative evaluation of emerging methods for errors of commission based on applications to the Davis-Besse (1985) event

    Energy Technology Data Exchange (ETDEWEB)

    Reer, B.; Dang, V.N.; Hirschberg, S. [Paul Scherrer Inst., Nuclear Energy and Safety Research Dept., CH-5232 Villigen PSI (Switzerland); Straeter, O. [Gesellschaft fur Anlagen- und Reaktorsicherheit (Germany)

    1999-12-01

    In considering the human role in accidents, the classical PSA methodology applied today focuses primarily on the omissions of actions required of the operators at specific points in the scenario models. A practical, proven methodology is not available for systematically identifying and analyzing the scenario contexts in which the operators might perform inappropriate actions that aggravate the scenario. As a result, typical PSA's do not comprehensively treat these actions, referred to as errors of commission (EOCs). This report presents the results of a joint project of the Paul Scherrer Institut (PSI, Villigen, Switzerland) and the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Garching, Germany) that examined some methods recently proposed for addressing the EOC issue. Five methods were investigated: 1 ) ATHEANA, 2) the Borssele screening methodology. 3) CREAM, 4) CAHR, and 5) CODA. In addition to a comparison of their scope, basic assumptions, and analytical approach, the methods were each applied in the analysis of PWR Loss of Feedwater scenarios based on the 1985 Davis-Besse event, in which the operator response included actions that can be categorized as EOCs. The aim was to compare how the methods consider a concrete scenario in which EOCs have in fact been observed. These case applications show how the methods are used in practical terms and constitute a common basis for comparing the methods and the insights that they provide. The identification of the potentially significant EOCs to be analysed in the PSA is currently the central problem for their treatment. The identification or search scheme has to consider an extensive set of potential actions that the operators may take. These actions may take place instead of required actions, for example, because the operators fail to assess the plant state correctly, or they may occur even when no action is required. As a result of this broad search space, most methodologies apply multiple schemes to

  14. Coordinated Control Scheme of Battery Energy Storage System (BESS) and Distributed Generations (DGs) for Electric Distribution Grid Operation

    DEFF Research Database (Denmark)

    Cha, Seung-Tae; Zhao, Haoran; Wu, Qiuwei

    2012-01-01

    into the islanding operation mode, while the centralized joint load frequency control (CJLFC) utilizing DGs handles the secondary frequency regulation. The BESS with the associated controllers has been modelled in Real-time digital simulator (RTDS) in order to identify the improvement of the frequency and voltage......This paper describes a coordinated control scheme of battery energy storage system (BESS) and distributed generations (DGs) for electric distribution grid operation. The BESS is designed to stabilize frequency and voltages as a primary control after the electric distribution system enters...... response. The modified IEEE 9-bus system, which is comprised of several DG units, wind power plant and the BESS, has been employed to illustrate the performance of the proposed coordinated flexible control scheme using RTDS in order to verify its practical efficacy....

  15. Areva's challenge for ''Georges Besse 2''

    International Nuclear Information System (INIS)

    Jemain, A.

    2003-01-01

    For its future uranium enrichment plant of its Tricastin site (Drome, France), the world nuclear leader Areva has abandoned the gaseous diffusion technique (of French origin) for the centrifugation technique, more economical and modular. This future plant, named 'Georges Besse 2' will require 3 billions of euros of investment and will supply a world market also estimated to 3 billions of euros and shared between Areva, Urenco (UK), Usec (US), Minatom (Russia), JNC (Japan) and CNNC (China). The first batches of enriched uranium will be produced using a thousand of centrifuges by 2007. (J.S.)

  16. Trending analysis of incidents involving primary water stress corrosion cracking on Alloy 600 components at U.S. PWRs

    International Nuclear Information System (INIS)

    Takahara, Shogo; Watanabe, Norio

    2006-01-01

    Primary Water Stress Corrosion Cracking (PWSCC) which occurs on Nickel based alloy (Alloy 600) is a worldwide concern since early 1980's. Recently several significant degradations that originate from PWSCC in the reactor coolant pressure boundary (RCPB) components have been observed at U.S. PWR plants (e.g. Oconee-3, Davis Besse). The United States Nuclear Regulation Commission (NRC) has issued generic communications to address this problem and, in response to the Davis Besse event in 2002, gave the inspection order EA-03-009 for the PWR licensees to implement the inspection of the reactor vessel heads depending upon the effective degradation years. As well, in Japan, PWSCC is considered one of the safety issues, in particular, for aged nuclear power plants and actually, some plants have experienced PWSCC on RCPB components. In the present study, we analyzed the U.S. experience with Alloy 600 degradation by reviewing the licensee event reports from 1999 to 2005 and examined the trend of them mainly focusing on affected components, characteristics of cracking and inspection approaches for detecting the PWSCC. This study indicates that PWSCC is found to be occurred on the RCPB components exposed to the environment with high temperature such as the reactor vessel head, and has the tendency to happen for specific manufactures and material according to the RCPB components. As well, it is shown that for several components, the non-destructive examination is generally needed to detect and/or confirm the PWSCC after the visual inspection and different repair techniques are applied depending on the components affected. (author)

  17. UF6 overfilling prevention at Eurodif production Georges Besse plant

    Energy Technology Data Exchange (ETDEWEB)

    Reneaud, J.M. [Eurodif Production, Pierrelatte (France)

    1991-12-31

    Risk of overfilling exists on different equipments of Georges BESSE Plant: cylinders, desublimers and intermediate tanks. The preventive measures are composed of technical devices: desublimers weighing, load monitoring alarms, automatic controls ... and procedures, training, safety organization. In thirteen years of operation, some incidents have occurred but none of them has caused any personal injuries. They are related and discussed. The main factors involved in the Sequoyah fuel facility accident on 1/4/1986 have been analyzed and taken into account.

  18. New Configuration and Novel Reclosing Procedure of Distribution System for Utilization of BESS as UPS in Smart Grid

    Directory of Open Access Journals (Sweden)

    Hun-Chul Seo

    2017-03-01

    Full Text Available This paper proposes a new configuration and novel reclosing procedure of a distribution system with a battery energy storage system (BESS used as an uninterruptible power supply (UPS in a smart grid. The proposed new configurations of the distribution systems are the installation of a circuit breaker (CB on both sides of the distribution line, the replacement of the recloser with a CB and protective relay, and the requirement of a communication method. The proposed reclosing procedure performs the reclosing of the CB at the load side and then judges the fault clearance using the load current. If the fault is cleared, the synchronism checking between the main source and the BESS is performed. After completing this, the CB at the main source side is reclosed. The smart grid environment, including a new distribution system, BESS, and reclosing method are modeled with the Electromagnetic Transients Program (EMTP/ATPDraw. To verify the proposed method, the various simulations according to the fault clearance time are performed and analyzed. The simulation results show that the BESS can be operated as a UPS and successful reclosing is possible.

  19. On-line monitoring of main coolant pump seals

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

    1984-06-01

    The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

  20. The genusGuenthera Andr. in Bess. (Brassicaceae, Brassiceae

    Directory of Open Access Journals (Sweden)

    Gómez-Campo, César

    2003-12-01

    Full Text Available A group of nine species -now included in Brassica— differ from all the other species in several characters, mainly in the stylar portion of their pistils always without seed primordia. Also in their branched subterranean stem (caudex with several leaf rosettes, their leaves entire to deeply pinnatifid but never pinnatisect, their shallowly notched cotyledons and their flattened, elliptic or ovoid seed contour. It is suggested to include these species under the generic denomination Guenthera Andr, in Bess. New ñames for the species and subspecies are provided, as well as a determination key for the species.Un grupo de nueve especies actualmente incluidas en Brassica difiere de todas las demás por varios caracteres, sobre todo por la porción estilar de sus pistilos, que siempre carece de primordios seminales. Además, por su tallo subterráneo ramificado, que forma un cáudex con varias rosetas; sus hojas de enteras hasta profundamente pinnatífidas, pero nunca pinnatisectas; sus cotiledones solo muy ligeramente escotados, y sus semillas, que tienden a ser elipsoidales o aplanadas. Se propone agruparlas todas bajo la denominación genérica Guenthera Andr, in Bess. Se detallan los nuevos nombres para las especies y Subespecies y se añade una clave para diferenciar las especies.

  1. Isotopic Survey of Lake Davis and the Local Groundwater

    Energy Technology Data Exchange (ETDEWEB)

    Ridley, M N; Moran, J E; Singleton, M J

    2007-08-21

    In September 2007, California Fish and Game (CAFG) plans to eradicate the northern pike from Lake Davis. As a result of the eradication treatment, local residents have concerns that the treatment might impact the local groundwater quality. To address the concerns of the residents, Lawrence Livermore National Laboratory (LLNL) recommended measuring the naturally occurring stable oxygen isotopes in local groundwater wells, Lake Davis, and the Lake Davis tributaries. The purpose of these measurements is to determine if the source of the local groundwater is either rain/snowmelt, Lake Davis/Big Grizzly Creek water or a mixture of Lake Davis/Big Grizzly Creek and rain/snowmelt. As a result of natural evaporation, Lake Davis and the water flowing into Big Grizzly Creek are naturally enriched in {sup 18}oxygen ({sup 18}O), and if a source of a well's water is Lake Davis or Big Grizzly Creek, the well water will contain a much higher concentration of {sup 18}O. This survey will allow for the identification of groundwater wells whose water source is Lake Davis or Big Grizzly Creek. The results of this survey will be useful in the development of a water-quality monitoring program for the upcoming Lake Davis treatment. LLNL analyzed 167 groundwater wells (Table 1), 12 monthly samples from Lake Davis (Table 2), 3 samples from Lake Davis tributaries (Table 2), and 8 Big Grizzly Creek samples (Table 2). Of the 167 groundwater wells sampled and analyzed, only 2 wells contained a significant component of evaporated water, with an isotope composition similar to Lake Davis water. The other 163 groundwater wells have isotope compositions which indicate that their water source is rain/snowmelt.

  2. Progress in Search for Antihelium with BESS

    Science.gov (United States)

    Sasaki, M.; Matsumoto, H.; Nozaki, M.; Saeki, T.; Abe, K.; Anraku, K.; Asoka, Y.; Fujikawa, M.; Fuke, H.; Imori, M.

    2002-01-01

    We have searched for antihelium nuclei in cosmic rays using the data obtained from balloon flights of the BESS magnetic spectrometer. The search was mainly based on track-quality selection, followed by rigidity analysis, and on the time-of-flight and dE/dx measurements by the scintillation counter hodoscope. We analysed all the data collected during 1993-2000 with a common analysis procedure. No antihelium nuclei events were found in the energy range from 1 to 14 GV. In order to determine a new upper limit, we have simulated the loss in the air and in the instrument of He (He-bar) using the GEANT/GHEISHA code. Combined with the data collected in 1993 through 2000, a new 95 % confidence upper limit for the ratio of He-bar/He at the top of the atmosphere of 6.8 x 10(exp -7) has been obtained to be after correcting for the interactions in the air and in the instruments.

  3. Measurements of Cosmic-Ray Proton and Helium Spectra from the BESS-Polar Long-Duration Balloon Flights Over Antarctica

    Science.gov (United States)

    Abe, K.; Fuke, H.; Haino, S.; Hams, T.; Hasegawa, M.; Horikoshi, A.; Itazaki, A.; Kim, K. C.; Kumazawa, T.; Kusumoto, A.; hide

    2016-01-01

    The BESS-Polar Collaboration measured the energy spectra of cosmic-ray protons and helium during two long-duration balloon flights over Antarctica in December 2004 and December 2007, at substantially different levels of solar modulation. Proton and helium spectra probe the origin and propagation history of cosmic rays in the galaxy, and are essential to calculations of the expected spectra of cosmic-ray antiprotons, positrons, and electrons from interactions of primary cosmic-ray nuclei with the interstellar gas, and to calculations of atmospheric muons and neutrinos. We report absolute spectra at the top of the atmosphere for cosmic-ray protons in the kinetic energy range 0.2-160 GeV and helium nuclei 0.15-80 GeV/nucleon. The corresponding magnetic rigidity ranges are 0.6-160 GV for protons and 1.1-160 GV for helium. These spectra are compared to measurements from previous BESS flights and from ATIC-2, PAMELA, and AMS-02. We also report the ratio of the proton and helium fluxes from 1.1 GV to 160 GV and compare to ratios from PAMELA and AMS-02.

  4. Obituary: Sumner P. Davis (1924-2008)

    Science.gov (United States)

    Feinberg, Jack

    2011-12-01

    University of California, Berkeley physicist Sumner P. Davis, a beloved teacher whose research centered on the optical spectroscopy of diatomic molecules found in the sun and other stars, died Dec. 31, 2008 in El Cerrito, CA after a brief illness. He was 84. After his military service during WWII, Davis finished his undergraduate work at UCLA in 1947, pursuing spectroscopy under the guidance of Joseph Ellis. Davis trained as a graduate student under molecular spectroscopist Francis Jenkins at UC Berkeley, where Davis used his ham radio expertise to construct an RF discharge to excite isotopes of diatomic selenium for his thesis. After receiving his Ph.D. from UC Berkeley, Davis went to MIT to postdoc under George Harrison, the premier artisan of finely-ruled diffraction gratings. In 1959, Jenkins invited Davis back to UC Berkeley to join the physics faculty, and Davis brought with him a highly prized gift - a diffraction grating presented to him by Harrison which Davis used for years to measure molecular spectra. At UC Berkeley Davis constructed a walk-in 15-foot-long spectrometer to produce detailed spectra of diatomic molecules of interest to astrophysics. With John G. Phillips he measured with high-precision the molecular constants of CN, C2, FeH, CS, SH and SiC2, TiO and others. Davis also studied the effect of the nuclear structure of Hg and Se on their optical spectra. He authored a book, Diffraction Grating Spectographs (1970), as well as monographs on CN and C2 spectra. Davis frequently traveled to the National Solar Observatory at Kitt Peak, to collect laboratory data using their Fourier transform spectrometer. He coauthored the book Fourier Transform Spectrometry (2001) with Mark C. Abrams and James Brault. In 1989, while returning to California after a long session on the spectrometer, his car, driven by Grace, his wife of 42 years, went off the road. Grace was killed but Sumner survived. Sumner Davis was, first and foremost, a consummate teacher

  5. Definable davies' theorem

    DEFF Research Database (Denmark)

    Törnquist, Asger Dag; Weiss, W.

    2009-01-01

    We prove the following descriptive set-theoretic analogue of a theorem of R. 0. Davies: Every σ function f:ℝ × ℝ → ℝ can be represented as a sum of rectangular Σ functions if and only if all reals are constructible.......We prove the following descriptive set-theoretic analogue of a theorem of R. 0. Davies: Every σ function f:ℝ × ℝ → ℝ can be represented as a sum of rectangular Σ functions if and only if all reals are constructible....

  6. 29 CFR 5.22 - Effect of the Davis-Bacon fringe benefits provisions.

    Science.gov (United States)

    2010-07-01

    ... 29 Labor 1 2010-07-01 2010-07-01 true Effect of the Davis-Bacon fringe benefits provisions. 5.22... Fringe Benefits Provisions of the Davis-Bacon Act § 5.22 Effect of the Davis-Bacon fringe benefits... paragraphs (a) and (b) of § 1.2 of this subtitle. The fringe benefits amendments enlarge the scope of this...

  7. The new enrichment plant of AREVA, a worthy heir of Georges Besse's industrial visions

    International Nuclear Information System (INIS)

    Oursel, L.

    2011-01-01

    The Georges Besse II enrichment plant was inaugurated on December 14., 2010. This plant was the most important investment in France during the last decade, about 3 billion euros. This plant is based on the centrifugation technology instead of the gaseous diffusion that is still in operation in the Georges Besse plant of EURODIF. This plant has been designed in an environment-friendly approach: the centrifugation technology uses 50 times less electricity than gaseous diffusion, does not require taking water from the Rhone river for cooling, does not produce sound nuisances, and the moderate height of the buildings allows a better integration in the environment. The low amount of matter involved in the centrifugation process gives the plant a high level of safety. The plant has a capacity of 7.5 millions UTS with a possible extension to 11 millions UTS. (A.C.)

  8. MEASUREMENTS OF COSMIC-RAY PROTON AND HELIUM SPECTRA FROM THE BESS-POLAR LONG-DURATION BALLOON FLIGHTS OVER ANTARCTICA

    Energy Technology Data Exchange (ETDEWEB)

    Abe, K.; Itazaki, A.; Kusumoto, A.; Matsukawa, Y.; Orito, R. [Kobe University, Kobe, Hyogo 657-8501 (Japan); Fuke, H. [Institute of Space and Astronautical Science, Japan Aerospace Exploration Agency (ISAS/JAXA), Sagamihara, Kanagawa 252-5210 (Japan); Haino, S.; Hasegawa, M.; Horikoshi, A.; Kumazawa, T.; Makida, Y.; Matsuda, S.; Matsumoto, K.; Nozaki, M. [High Energy Accelerator Research Organization (KEK), Tsukuba, Ibaraki 305-0801 (Japan); Hams, T.; Mitchell, J. W. [NASA-Goddard Space Flight Center (NASA-GSFC), Greenbelt, MD 20771 (United States); Kim, K. C.; Lee, M. H.; Myers, Z. [IPST, University of Maryland, College Park, MD 20742 (United States); Nishimura, J., E-mail: Kenichi.Sakai@nasa.gov [The University of Tokyo, Bunkyo, Tokyo 113-0033 (Japan); and others

    2016-05-10

    The BESS-Polar Collaboration measured the energy spectra of cosmic-ray protons and helium during two long-duration balloon flights over Antarctica in 2004 December and 2007 December at substantially different levels of solar modulation. Proton and helium spectra probe the origin and propagation history of cosmic rays in the galaxy, and are essential to calculations of the expected spectra of cosmic-ray antiprotons, positrons, and electrons from interactions of primary cosmic-ray nuclei with the interstellar gas, and to calculations of atmospheric muons and neutrinos. We report absolute spectra at the top of the atmosphere for cosmic-ray protons in the kinetic energy range 0.2–160 GeV and helium nuclei in the range 0.15–80 GeV/nucleon. The corresponding magnetic-rigidity ranges are 0.6–160 GV for protons and 1.1–160 GV for helium. These spectra are compared to measurements from previous BESS flights and from ATIC-2, PAMELA, and AMS-02. We also report the ratio of the proton and helium fluxes from 1.1 to 160 GV and compare this to the ratios from PAMELA and AMS-02.

  9. Two-Stage Battery Energy Storage System (BESS in AC Microgrids with Balanced State-of-Charge and Guaranteed Small-Signal Stability

    Directory of Open Access Journals (Sweden)

    Bing Xie

    2018-02-01

    Full Text Available In this paper, a two-stage battery energy storage system (BESS is implemented to enhance the operation condition of conventional battery storage systems in a microgrid. Particularly, the designed BESS is composed of two stages, i.e., Stage I: integration of dispersed energy storage units (ESUs using parallel DC/DC converters, and Stage II: aggregated ESUs in grid-connected operation. Different from a conventional BESS consisting of a battery management system (BMS and power conditioning system (PCS, the developed two-stage architecture enables additional operation and control flexibility in balancing the state-of-charge (SoC of each ESU and ensures the guaranteed small-signal stability, especially in extremely weak grid conditions. The above benefits are achieved by separating the control functions between the two stages. In Stage I, a localized power sharing scheme based on the SoC of each particular ESU is developed to manage the SoC and avoid over-charge or over-discharge issues; on the other hand, in Stage II, an additional virtual impedance loop is implemented in the grid-interactive DC/AC inverters to enhance the stability margin with multiple parallel-connected inverters integrating at the point of common coupling (PCC simultaneously. A simulation model based on MATLAB/Simulink is established, and simulation results verify the effectiveness of the proposed BESS architecture and the corresponding control diagram.

  10. Environmental assessment overview, Davis Canyon site, Utah

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considered for the first repository. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of five sites suitable for characterization. 3 figs

  11. Environmental assessment: Davis Canyon site, Utah

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EA. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains one other potentially acceptable site -- the Lavender Canyon site. Although the Lavender Canyon site is suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. Furthermore, the DOE has fond that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of five sites suitable for characterization. 181 figs., 175 tabs

  12. Environmental assessment: Davis Canyon site, Utah

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high- level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EA. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains one other potentially acceptable site -- the Lavender Canyon site. Although the Lavender Canyon site is suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. Furthermore, the DOE has found that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of the five sites suitable for characterization.

  13. Environmental assessment: Davis Canyon site, Utah

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EA. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains one other potentially acceptable site -- the Lavender Canyon site. Although the Lavender Canyon site is suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. Furthermore, the DOE has fond that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of five sites suitable for characterization. 181 figs., 175 tabs.

  14. Environmental assessment: Davis Canyon site, Utah

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EA. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considering for the first repository. This setting contains one other potentially acceptable site -- the Lavender Canyon site. Although the Lavender Canyon site is suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. Furthermore, the DOE has found that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of five sites suitable for characterization.

  15. Environmental assessment: Davis Canyon site, Utah

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EA. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considering for the first repository. This setting contains one other potentially acceptable site -- the Lavender Canyon site. Although the Lavender Canyon site is suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. Furthermore, the DOE has found that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of five sites suitable for characterization

  16. Environmental assessment: Davis Canyon site, Utah

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high- level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EA. The Davis Canyon site is in the Paradox Basin, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains one other potentially acceptable site -- the Lavender Canyon site. Although the Lavender Canyon site is suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. On the basis of the evaluations reported in this EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. Furthermore, the DOE has found that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Davis Canyon site as one of the five sites suitable for characterization

  17. Experiences concerning reactor pressure vessel head penetration inspections; Erfahrungen mit Pruefungen von Reaktordruckbehaelter-Deckeldurchfuehrungen

    Energy Technology Data Exchange (ETDEWEB)

    Debnar, Angelika [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2009-07-01

    Globally observed damage at the control rod drive mechanism nozzles in PWR-type reactors (Bugey-3, Oconee 1,2,3 and ANO-1, David Besse) have triggered enhanced inspection of reactor pressure vessel (RPV) head penetrations. In Germany the regulations require a periodic inspection especially of dissimilar welds. Westinghouse has developed an automated measuring system for RPV heads aimed to inspect welded joints at open nozzles of nozzles with thermosleeves. The testing technology with remote controlled robotics is supposed to perform a weld inspection as complete as possible, restraints result from constructive difficulties for the accessibility. The new gap-scanner DE2008 was qualified at the mock-up and was implemented into the periodic in-service inspection of Neckarwestheim-1.

  18. Achievements of the IAEA technical working group on life management of nuclear power plants (TWG-LMNPP) under the chairmanship of Acad. Myrddin Davies

    International Nuclear Information System (INIS)

    Kang, K.-S.; Tipping, Philip

    2004-01-01

    This meeting, organised by CRISM-PROMETEY in St Petersburg, Russia, is held to honour the memory of Academician Myrddin Davies, who passed away due to a tragic road accident on 11 March 2003 in Stretton, England. Academician Myrddin Davies started technical collaboration with the IAEA in the early 1980s, and in 1985 became chairman of the International Working Group on Reliability of Reactor Pressure Components (IWG-RRPC). Under his chairmanship this grew to become the Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP) covering broader issues and with world wide collaboration. An insight to the creation, working methods and achievements of the TWG-LMNPP is given in this paper. Acad. Myrddin Davies was a competent chairman at many specialist meetings, major conferences hosted by IAEA, other European organizations and Nuclear Engineering International activities. The direction given to the TWG-LMNPP by Acad. Myrddin Davies is shown to have made a significant contribution to the safe use of nuclear energy. Major contributions to nuclear technology of the TWG-LMNPP, during the Chairmanship of Myrddin Davies, are thus cited

  19. Safety significance of inadvertent operation of motor-operated valves in nuclear power plants

    International Nuclear Information System (INIS)

    Ruger, C.J.; Higgins, J.C.; Carbonaro, J.F.; Hall, R.E.

    1994-01-01

    Concerns about the consequences of valve mispositioning were brought to the forefront following an event at Davis Besse in 1985. The concern related to the ability to reposition open-quotes position-changeableclose quotes motor-operated valves (MOVs) from the control room in the event of their inadvertent operation and was documented in U.S. Nuclear Regulatory Commission (USNRC) Bulletin 85-03 and Generic Letter (GL) 89-10. The mispositioned MOVs may not be able to be returned to their required position due to high differential pressure or high flow conditions across the valves. The inability to reposition such valves may have significantly safety consequences, as in the Davis Besse event. However, full consideration of such mispositioning in safety analyses and in MOV test programs can be labor intensive and expensive. Industry raised concerns that consideration of position-changeable valves under GL 89-10 would not decrease the probability of core damage to an extent that would justify licensee costs. As a response, Brookhaven National Laboratory has conducted separate scoping studies for both boiling water reactors (BWRs) and pressurized water reactors (PWRs) using probabilistic risk assessment (PRA) techniques to determine if such valve mispositioning by itself is significant to safety. The approach used internal events PRA models to survey the order of magnitude of the risk-significance of valve mispositioning by considering the failure of selected position-changeable MOVs. The change in core damage frequency was determined for each valve considered, and the results were presented as a risk increase ratio for each of four assumed MOV failure rates. The risk increase ratios resulting from this failure rate sensitivity study can be used as a basis for a determination of the risk-significance of the MOV mispositioning issues for BWRs and PWRs

  20. Obituary: Leverett Davis, Jr., 1914-2003

    Science.gov (United States)

    Jokipii, Jack Randolph

    2004-12-01

    of the heliosphere is not yet known for certain, but is certainly greater than some 90 AU (the current distance to the Voyager 1 spacecraft), and probably of the order of, or perhaps greater than, 100 AU. Observational and theoretical investigations of the boundary of this cavity are currently a very active area of research. In the 1960's, Davis's interests in astrophysical and solar magnetic fields, energetic particles and plasmas led naturally to investigations in the new field of space physics, where observations from spacecraft were revolutionizing our understanding. He wrote important papers unraveling the basic physical processes governing the motions of trapped particles in the radiation belts of planets such as the Earth and Jupiter. This work led naturally to his deep involvement in the early space program, where detailed in situ observations of these phenomena became possible. He became a consultant to one of the early companies developing spacecraft, and this led to a number of pioneering contributions to our understanding of interplanetary space. Davis was one of the true pioneers in the exploration of the plasmas and their associated magnetic fields in space using in situ observations from spacecraft, which began in the late 1950's. He participated effectively as a co-investigator in several of the early planetary missions to Venus in 1962 (Mariner 2), to Mars in 1964 (Mariner4), to Jupiter in 1973-74 (Pioneer10, 11) and to Saturn in 1989 (Pioneer11). The Pioneer spacecraft returned data for nearly 30 years, until the last signal was received from Pioneer 10 in 2002. He continued working on spacecraft data until the early 1980's when he retired. In both his personal and professional life, Davis was a man of very high standards and great personal integrity. He was a devoted family man who enjoyed nothing more than a road trip including camping, with his family. He was serious about his work and responsibilities, but also had a subtly infectious sense of

  1. Nuclear Regulatory Commission Issuances, August 1981

    International Nuclear Information System (INIS)

    1981-01-01

    Contents include: Issuances of the Nuclear Regulatory Commission--Metropolitan Edison Company (Three Mile Island Nuclear Station, Unit No. 1), Metropolitan Edison Company, et al. (Three Mile Island Nuclear Station, Unit 1), Westinghouse Electric Corp. (Export of LEU to the Philippines); Issuances of Atomic Safety and Licensing Appeal Boards--Duke Power Company (Amendment to Materials License SNM-1773--Transportation of Spent Fuel from Oconee Nuclear Station for Storage at McGuire Nuclear Station); Issuances of the Atomic Safety Licensing Boards--Commonwealth Edison Company (Byron Station, Units 1 and 2), Dairyland Power Cooperative (La Crosse Boiling Water Reactor, Operating License and Show Cause), Florida Power and Light Company (St. Lucie Plant, Unit No. 2), Florida Power and Light Company (Turkey Point Nuclear Generating, Units 3 and 4), Metropolitan Edison Company (Three Mile Island Nuclear Station, Unit 1) Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), The Regents of the University of California (UCLA Research Reactor), The Toledo Edison Company, et al. (Davis-Besse Nuclear Power Station, Units 2 and 3: Terminiation of Proceedings); Issuances of the Directors Denial--Florida Power and Light Company

  2. SoC-Based Output Voltage Control for BESS with a Lithium-Ion Battery in a Stand-Alone DC Microgrid

    Directory of Open Access Journals (Sweden)

    Seung-Yeong Yu

    2016-11-01

    Full Text Available This paper proposes a new DC output voltage control for a battery energy storage system (BESS with a lithium-ion battery based on the state of charge (SoC. The proposed control scheme was verified through computer simulations for a typical stand-alone DC microgrid, which consists of a BESS, photovoltaic (PV panel, engine generator (EG, and DC load. A scaled hardware prototype for a stand-alone DC microgrid was set up in the lab, in which the proposed control scheme was loaded in a DSP controller. The experimental results were compared with the simulation results for performance verification. The proposed control scheme provides relatively lower variation of the DC grid voltage than the conventional droop control.

  3. The Davis Strait

    DEFF Research Database (Denmark)

    The Bureau of Minerals and Petroleum (BMP) is planning for further exclusive licences for exploration and exploitation of hydrocarbons in the Greenland off shore areas of Davis Strait. To support the decision process BMP has asked DCE - Danish Centre for Environment and Energy and the Greenland I...

  4. 27 CFR 9.155 - Texas Davis Mountains.

    Science.gov (United States)

    2010-04-01

    ...) “Fort Davis, Texas,” 1985. (2) “Mount Livermore, Texas—Chihuahua,” 1985. (c) Boundary. The Texas Davis... follows Highway 166 in a southwesterly direction onto the Mt. Livermore, Texas-Chihuahua, U.S.G.S. map; (6... Grapevine Canyon on the Mt. Livermore, Texas-Chihuahua, U.S.G.S. map; (14) The boundary then proceeds in a...

  5. Optimal Operation and Management of Smart Grid System with LPC and BESS in Fault Conditions

    Directory of Open Access Journals (Sweden)

    Ryuto Shigenobu

    2016-12-01

    Full Text Available Distributed generators (DG using renewable energy sources (RESs have been attracting special attention within distribution systems. However, a large amount of DG penetration causes voltage deviation and reverse power flow in the smart grid. Therefore, the smart grid needs a solution for voltage control, power flow control and power outage prevention. This paper proposes a decision technique of optimal reference scheduling for a battery energy storage system (BESS, inverters interfacing with a DG and voltage control devices for optimal operation. Moreover, the reconfiguration of the distribution system is made possible by the installation of a loop power flow controller (LPC. Two separate simulations are provided to maintain the reliability in the stable power supply and economical aspects. First, the effectiveness of the smart grid with installed BESS or LPC devices is demonstrated in fault situations. Second, the active smart grid using LCPs is proposed. Real-time techniques of the dual scheduling algorithm are applied to the system. The aforementioned control objective is formulated and solved using the particle swarm optimization (PSO algorithm with an adaptive inertia weight (AIW function. The effectiveness of the optimal operation in ordinal and fault situations is verified by numerical simulations.

  6. An overview on emerging bioelectrochemical systems (BESs): Technology for sustainable electricity, waste remediation, resource recovery, chemical production and beyond

    NARCIS (Netherlands)

    Bajracharya, S.; Sharma, M.; Mohanakrishna, Gunda; Benneton, Xochitl Dominguez; Strik, D.P.B.T.B.; Sarma, Priyangshu M.; Pant, Deepak

    2016-01-01

    Bioelectrochemical systems (BESs) are unique systems capable of converting chemical energy into electrical energy (and vice-versa) while employing microbes as catalysts. Such organic wastes including low-strength wastewaters and lignocellulosic biomass were converted into electricity with microbial

  7. Knowledge Brokerage for Environmentally Sustainable Sanitation. Position Paper and Guidelines from the EU-FP7 BESSE project.

    NARCIS (Netherlands)

    BESSE, Project team; Bijker, W.E.; Caiati, Giovanni; d'Andrea, Luciano

    2012-01-01

    The EU-funded BESSE project explores how sanitation in Europe can be made more sustainable. European sanitation is still based on 19th and early 20th century technologies and management systems. These systems do not adequately respond to the sustainable development needs of the 21st century, such as

  8. 75 FR 57299 - First Energy Nuclear Operating Company; Notice of Receipt and Availability of Application for...

    Science.gov (United States)

    2010-09-20

    ... NUCLEAR REGULATORY COMMISSION [NRC-2010-0298; Docket No. 50-346] First Energy Nuclear Operating Company; Notice of Receipt and Availability of Application for Renewal of Davis Besse Nuclear Power...Energy Nuclear Operating Company, filed pursuant to Section 104(b) of the Atomic Energy Act of 1954, as...

  9. Sexual maturity cycle and spawning of Greenland halibut Reinhardtius hippoglossoides in the Davis Strait

    DEFF Research Database (Denmark)

    Gundersen, A. C.; Stenberg, Claus; Fossen, I.

    2010-01-01

    Female sexual maturation cycle and the main spawning time of Greenland halibut Reinhardtius hippoglossoides in the Davis Strait were studied through regularly collected samples during 1 year starting in spring 2003. Samples were collected from the southern slope of the Davis Strait Ridge between...

  10. Douglas Davis / Douglas Davis ; interv. Tilman Baumgärtel

    Index Scriptorium Estoniae

    Davis, Douglas

    2006-01-01

    Ameerika kunstnikust Douglas Davisest (sünd. 1933) ja tema loonmingust, intervjuu kunstnikuga 8. 05. 1999 Osnabrückis. D. Davis oma interaktiivsetest performance'itest "Austrian Tapes" ja "Florence Tapes" (1970-ndad), Interneti-projektist "Terrible Beauty", sateliidiperformance'ist "Seven Thoughts" (1976), teleperformance'ist "The Last Nine Minutes" (1977), Vitali Komari ja Aleksander Melamidiga koos tehtud projektist "Questions Moscow New York" (1975-1976), võrguprojektidest "The World's Longest Sentence" (1994, asub New Yorgi Whitney Muuseumis), "MetaBody" jm.

  11. Kuidas kirjutatakse ajalugu? / Natalie Zemon Davis ; interv. Marek Tamm

    Index Scriptorium Estoniae

    Davis, Natalie Zemon

    2007-01-01

    Ülevaade Princentoni Ülikooli emeriitprofessori ja Ameerika Ajalooühingu endise presidendi N. Z. Davis'e teostest. Varem ilm.: Ajalugu, filmikunst ja heidikud : intervjuu Natalie Zemon Davisega // Davis, Natalie Zemon. Martin Guerre'i tagasitulek. - Tallinn, 2002. - Lk. 162-177

  12. High cost of nuclear power plants

    International Nuclear Information System (INIS)

    Bassett, C.

    1978-01-01

    Retroactive safety standards were found to account for over half the costs of a nuclear power plant and point up the need for an effective cost-benefit analysis of changes made by the Nuclear Regulatory Commission after construction has started. The author compared the Davis-Besse Unit No. 1 construction-cost estimates with the final-cost increases during a rate-case investigation in Ohio. He presents data furnished for ten of the largest construction contracts to illustrate the cost increases involving fixed hardware and intensive labor. The situation was found to repeat with other utilities across the country even though safeguards against irresponsible low bidding were introduced. Low bidding was found to continue, encouraged by the need for retrofitting to meet regulation changes. The average cost per kilowatt of major light-water reactors is shown to have increased from $171 in 1970 to $555 in 1977, while construction duration increased from 43.4 to 95.6 months during the same period

  13. French Regulatory Framework for the Recycling/Reuse of Nuclear Waste and the Dismantling of George Besse Gaseous Diffusion Plant

    Energy Technology Data Exchange (ETDEWEB)

    Themines, R., E-mail: robert.themines@areva.com [AREVA (France)

    2011-07-15

    The regulatory framework in France governing the management of materials containing low levels of radionuclides is described. The plans for the management of the materials arising from the dismantling of the Georges Besse Gaseous Diffusion Plant are described as an example of the application of the regulations. (author)

  14. Davis Meeting on Cosmic Inflation

    CERN Document Server

    Kaloper, N; Knox, L; Cosmic Inflation

    2003-01-01

    The Davis Meeting on Cosmic Inflation marked an exciting milestone on the road to precision cosmology. This is the index page for the proceedings of the conference. Individual proceedings contributions, when they appear on this archive, are linked from this page.

  15. Davis Strædet

    DEFF Research Database (Denmark)

    Råstofdirektoratet planlægger at udbyde flere licensområder med henblik på efterforskning og udvinding af olie og gas i den grønlandske del af Davis Stræde. Som en del af beslutningsgrundlaget har Råstofdirektoratet bedt DCE – Nationalt Center for Miljø og Energi og Grønlands Naturinstitut om at ...

  16. Report on the emergency evacuation review team on emergency response plans for the Perry and Davis-Besse nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This book is a report by Ohio's Emergency Evacuation Review Team, at the request of Governor Richard Celeste. The Team concludes that the current emergency response plan for Ohio's reactors is inadequate to protect the public and recommends changes in the current emergency plant requirements. The report also includes a summary of the litigation that has occurred since Celeste withdrew his support for the plans, a list of experts consulted, and sources used to prepare the report. An important document, and a study which every state should undertake

  17. [Sir Humphry Davy, the discoverer of anesthetic action of nitrous oxide--Davy and poets of British Romanticism and inhalation of laughing gas by his friends].

    Science.gov (United States)

    Fujita, T

    1998-01-01

    In "Dove Cottage", the old house of the poet laureate William Wordsworth (1770-1850) in Grasmere, England, there is a portrait of Sir Humphry Davy (1778-1829). In 1804, Wordsworth invited his young friend to his home. Davy's works in the field of chemistry are well known. Interestingly enough, once he wished he could be a poet. His future seemed to be prosperous and delightful. He was highly evaluated by Robert Southey, poet laureate. But he has chosen the way of chemist. The author found some facts from literatures and received some information by courtesy of the Wordsworth Trust, Centre for British Romanticism. Davy's life and his works were introduced chronologically.

  18. 40 Years in Applied Linguistics: An Interview with Alan Davies

    Science.gov (United States)

    Kunnan, Antony John

    2005-01-01

    This article presents an interview with Professor Alan Davies who was born in Wales, studied at Oxford University and Birmingham University, and taught in Scotland at the University of Edinburgh, completing 40 years this year. Professor Davies has travelled widely to give invited talks and seminars, participate in applied linguistics conferences,…

  19. Comparison of licensing activities for operating plants designed by Babcock and Wilcox

    International Nuclear Information System (INIS)

    Thoma, J.O.

    1985-01-01

    This report provides a comparison of a number of licensing activities for the operating Babcock and Wilcox (B and W) plants with emphasis on Rancho Seco. The factors selected were a comparison of staff resources expended in FY84, active licensing action reviews, implementation of NUREG-0737 modifications, exemptions to regulations, SALP reports, enforcement actions, and Licensee Event Reports (LERs). The eight licensed operating plants examined are as follows: Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3, Davis Besse, Oconee Units 1, 2, and 3, Rancho Seco, and Three Mile Island Unit 1 (TMI-1)

  20. Semi-automated uranium analysis by a modified Davies--Gray procedure

    International Nuclear Information System (INIS)

    Swanson, G.C.

    1977-01-01

    To rapidly and reliably determine uranium in fuel materials a semi-automated implementation of the Davies-Gray uranium titration was developed. The Davies-Gray method is essentially a three step procedure. First uranium is reduced quantitatively from +6 valence to +4 valence by excess of iron (II) in strong phosphoric acid in the absence of nitrite. Prior to the uranium reduction nitrite is destroyed by addition of sulfamic acid. In the second step iron (II) is selectively oxidized to iron (III) by nitric acid in the presence of Mo (VI) catalyst. Finally after dilution to reduce phosphate concentration, the uranium is titrated to U (VI) by standard dichromate. The original sluggish colorimetric endpoint determination used by Davies and Gray is seldom used since New Brunswick Laboratory discovered that addition of vanadium (IV) just prior to titration sufficiently improves reaction rate to allow a potentiometric endpoint determination. One of the advantages of the Davies-Gray uranium titration is that it is quite specific for uranium, most common impurity elements do not interfere with the analysis, and specifically high levels of Pu, Th, and Fe are tolerated

  1. The UC Davis/NIH NeuroMab Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The mission of the UC Davis/NIH NeuroMab facility is to generate and distribute high quality, validated mouse monoclonal antibodies against molecular targets found...

  2. USA suursaadik : toetame Eestit / Stanley Davis Phillips ; interv. Erkki Bahovski

    Index Scriptorium Estoniae

    Phillips, Stanley Davis

    2007-01-01

    USA suursaadik Eestis Stanley Davis Phillips vastab küsimustele, mis puudutavad USA positsiooni Tõnismäe pronkssõduri suhtes, Eesti saatkonna piiramist Moskvas, USA ja Venemaa suhteid ning koostööd, sõda terrorismiga, USA kava paigutada Tšehhi ja Poolasse raketitõrjebaasid, Eesti presidendi Toomas Hendrik Ilvese visiiti USAsse. Lisa: Stanley Davis Phillips. Ilmunud ka: Postimees : na russkom jazõke 16. mai lk. 5

  3. 75 FR 80549 - FirstEnergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Exemption

    Science.gov (United States)

    2010-12-22

    ... process for determination of RT PTS , the reference temperature RT NDT , evaluated for the end of license...]F (the weld wire heat-specific chemical composition, via the methodology of RG 1.99, Revision 2... requirements. Based on the above, no new accident precursors are created by allowing an exemption to use an...

  4. 48 CFR 970.2204-1-1 - Administrative controls and criteria for application of the Davis-Bacon Act in operational or...

    Science.gov (United States)

    2010-10-01

    ... 48 Federal Acquisition Regulations System 5 2010-10-01 2010-10-01 false Administrative controls... Administrative controls and criteria for application of the Davis-Bacon Act in operational or maintenance.... The proving out of investigative findings and theories of a scientific and technical nature may...

  5. Lööme kaasa! / Stanley Davis Phillips

    Index Scriptorium Estoniae

    Phillips, Stanley Davis

    2008-01-01

    USA Eesti-suursaadik Stanley Davis Phillips ütleb, et nad on saatkonnas moodustanud meeskonna, mis koosneb ameeriklastest kui ka eestlastest, et liituda üleriigilise ebaseaduslike prügilate likvideerimise üritusega "Teeme ära 2008"

  6. Historian Norman Davies Receives Estonian State Decoration

    Index Scriptorium Estoniae

    2008-01-01

    Eesti suursaadik Suurbritannias Margus Laidre andis 15. märtsil 2008 Londonis Suurbritannia ajaloolasele Ivor Norman Richard Davies'ele üle talle president Toomas Hendrik Ilvese poolt annetatud Maarjamaa Risti III klassi teenetemärgi

  7. Fort Davis National Historic Site : acoustical monitoring

    Science.gov (United States)

    2013-06-01

    During the summer of 2010 (September - October 2010), the Volpe Center collected baseline acoustical data at Fort Davis National Historic Site (FODA)at two sites deployed for approximately 30 days each. The baseline data collected during this period ...

  8. Draft environmental assessment: Davis Canyon site, Utah. Nuclear Waste Policy Act (Section 112)

    International Nuclear Information System (INIS)

    1984-12-01

    In February 1983, the US Department of Energy (DOE) identified the Davis Canyon site in Utah, as one of nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the Davis Canyon site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for Nuclear Waste Repositories. These evaluations are reported in this draft environmental assessment (EA), which is being issued for public review and comment. The DOE findings and determinations that are based on these evaluations are preliminary and subject to public review and comment. A final EA will be prepared after considering the comments received. On the basis of the evaluations reported in this draft EA, the DOE has found that the Davis Canyon site is not disqualified under the guidelines. The site is in the Paradox Basin, which is one of five distinct geohydrologic settings considered for the first repository. This setting contains one other potentially acceptable site - the Lavender Canyon site. Although the Lavender Canyon site appears to be suitable for site characterization, the DOE has concluded that the Davis Canyon site is the preferred site in the Paradox Basin. Furthermore, the DOE finds that the site is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is proposing to nominate the Davis Canyon site as one of five sites suitable for characterization. Having compared the Davis Canyon site with the other four sites proposed for nomination, the DOE has determined that the Davis Canyon site is not one of the three preferred sites for recommendation to the President as candidates for characterization

  9. A Hydrodynamic Study of Davis Pond, Near New Orleans, LA

    Science.gov (United States)

    2008-08-01

    District 2004). This project will make the Barataria estuary a more prolific producer of oysters, shrimp, crab , and fish, as well as a major habitat...yards 0.9144 meters ERDC/CHL TR-08-11 1 1 Introduction The Davis Pond freshwater diversion project is a salinity -control structure located in St...Mathematical model of estuarial sediment transport. Technical Report D-77-12. Vicksburg, MS: U.S. Army Engineer Waterways Experiment Station

  10. Unsteady flow analysis of combustion processes in a Davis gun

    Energy Technology Data Exchange (ETDEWEB)

    Cho, H.-C.; Shin, H.D. [Korea Advanced Inst. of Science and Technology, Mechanical Engineering Dept., Taejon (Korea, Republic of); Yoon, J.-K. [Hansung Univ., School of Industrial and System Engineering, Seoul (Korea, Republic of)

    1999-09-01

    The Davis gun, a type of recoilless gun, had the advantages of requiring less rear area and less powder than a conventional recoilless gun. The unsteady pressure and flow fields of a Davis gun were numerically simulated by using a two-phase fluid dynamic model. Numerical simulation results were compared with experimental values to evaluate the feasibility of the interior ballistic model. The interior ballistics in a Davis gun with a simple countermass were predicted with the computational model. It was shown that the pressure-time curves matched well between experimental data and numerical analysis except in the vicinity of the peak pressure and steep pressure gradient. The predicted muzzle velocity of projectile and countermass was closely similar to the experimental one. In this study, large pressure waves were not observed since the initial porosity was relatively high ({phi}{sub 0}0.867) and the charge was ignited at the centre of the granular bed. (Author)

  11. Removal of heavy metals from fly ash leachate using combined bioelectrochemical systems and electrolysis

    International Nuclear Information System (INIS)

    Tao, Hu-Chun; Lei, Tao; Shi, Gang; Sun, Xiao-Nan; Wei, Xue-Yan; Zhang, Li-Juan; Wu, Wei-Min

    2014-01-01

    Highlights: • Heavy metals removal from MSWI fly ash with BES and electrolysis was confirmed. • 98.5% of Cu(II), 95.4% of Zn(II) and 98.1% of Pb(II) removal were achieved in reactors. • BESs can remove some heavy metals in fly ash with energy saving. -- Abstract: Based on environmental and energetic analysis, a novel combined approach using bioelectrochemical systems (BES) followed by electrolysis reactors (ER) was tested for heavy metals removal from fly ash leachate, which contained high detectable levels of Zn, Pb and Cu according to X-ray diffraction analysis. Acetic acid was used as the fly ash leaching agent and tested under various leaching conditions. A favorable condition for the leaching process was identified to be liquid/solid ratio of 14:1 (w/w) and leaching duration 10 h at initial pH 1.0. It was confirmed that the removal of heavy metals from fly ash leachate with the combination of BESs and ER is feasible. The metal removal efficiency was achieved at 98.5%, 95.4% and 98.1% for Cu(II), Zn(II), and Pb(II), respectively. Results of scanning electron microscopy (SEM) and energy-dispersive X-ray spectroscopy (EDS) indicated that Cu(II) was reduced and recovered mainly as metal Cu on cathodes related to power production, while Zn(II) and Pb(II) were not spontaneously reduced in BESs without applied voltage and basically electrolyzed in the electrolysis reactors

  12. Norman Davies : kogu ajaloolise tõe kirjapanek pole paratamatult võimalik / Norman Davies ; interv. Marek Tamm

    Index Scriptorium Estoniae

    Davies, Norman, 1939-

    2008-01-01

    Intervjuu Läänes Ida-Euroopa ajalugu tutvustanud Briti ajaloolasega. Vt. samas: Bahovski, Erkki. Puhas ajalugu? Tartu Ülikooli aulas 8. oktoobril 2008 "Presidendi Kärajate" raames toimunud president Toomas Hendrik Ilvese ja Oxfordi ülikooli ajalooprofessori Norman Daviese avalikust kahekõnest. Vestlust juhatas Eesti suursaadik Suurbritannias Margus Laidre. Autori sõnul tahtis Eesti president rääkida tänapäeva poliitikast, N. Davies aga ajaloost

  13. Seeing your way to health: the visual pedagogy of Bess Mensendieck's physical culture system.

    Science.gov (United States)

    Veder, Robin

    2011-01-01

    This essay examines the images and looking practices central to Bess M. Mensendieck's (c.1866-1959) 'functional exercise' system, as documented in physical culture treatises published in Germany and the United States between 1906 and 1937. Believing that muscular realignment could not occur without seeing how the body worked, Mensendieck taught adult non-athletes to see skeletal alignment and muscular movement in their own and others' bodies. Three levels of looking practices are examined: didactic sequences; penetrating inspection and appreciation of physiological structures; and ideokinetic visual metaphors for guiding movement. With these techniques, Mensendieck's work bridged the body cultures of German Nacktkultur (nudism), American labour efficiency and the emerging physical education profession. This case study demonstrates how sport historians could expand their analyses to include practices of looking as well as questions of visual representation.

  14. Liquid jet experiments: relevance to inertial confinement fusion reactors

    International Nuclear Information System (INIS)

    Hoffman, M.A.

    1981-01-01

    In order to try to find a reactor design which offered protection against neutron damage, studies were undertaken at LLNL (the Lawrence Livermore National Laboratory) of self-healing, renewable liquid-wall reactor concepts. In conjuction with these studies, were done a seris of small-scale aer jet experiments were done over the past several years at UCD (University of California, Davis Campus) to simulate the behavior of liquid lithium (or lithium-lead) jets in these liquid-wall fusion reactor concepts. Extropolating the results of these small-scale experiments to the large-scale lithium jets, tentatively concluded that the lithium jet can be re-established after the microexplosion, and with careful design the jets should not breakup due to instabilities during the relatively quiscent period between MICROEXPLOSIONS

  15. Basement control of alkalic flood rhyolite magmatism of the Davis Mountains volcanic field, Trans-Pecos Texas, U.S.A.

    Science.gov (United States)

    Parker, Don F.; White, John C.; Ren, Minghua; Barnes, Melanie

    2017-11-01

    Voluminous silicic lava flows, erupted 37.4 Ma from widespread centers within the Davis Mountains Volcanic Field (DMVF), covered approximately 10,000 km2 with an initial volume as great as 1000 km3. Lava flows form three major stratigraphic units: the Star Mountain Rhyolite (minimum 220 km3) of the eastern Davis Mountains and adjacent Barilla Mountains, the Crossen Formation ( 75 km3) of the southern Davis Mountains, and the Bracks Rhyolite ( 75 km3) of the Rim Rock region west of the Davis Mountains proper. Similar extensive rhyolite lava also occurs in slightly younger units (Adobe Canyon Rhyolite, 125 km3, 37.1 Ma), Sheep Pasture Formation ( 125 km3, 36 Ma) and, less voluminously, in the Paisano central volcano ( 36.9 Ma) and younger units in the Davis Mountains. Individual lava flows from these units formed fields as extensive as 55 km and 300-m-thick. Flood rhyolite lavas of the Davis Mountains are marginally peralkaline quartz trachyte to low-silica rhyolite. Phenocrysts include alkali feldspar, clinopyroxene, FeTi oxides, and apatite, and, rarely, fayalite, as well as zircon in less peralkaline units. Many Star Mountain flows may be assigned to one of four geochemical groupings. Temperatures were moderately high, ranging from 911 to 860 °C in quartz trachyte and low silica rhyolite. We suggest that flood rhyolite magma evolved from trachyte magma by filter pressing processes, and trachyte from mafic magma in deeper seated plutons. The Davis Mountains segment of Trans-Pecos Texas overlies Grenville basement and is separated from the older Southern Granite and Rhyolite Province to the north by the Grenville Front, and from the younger Coahuila terrane to the south by the Ouachita Front. We suggest that basement structure strongly influenced the timing and nature of Trans-Pecos magmatism, probably in varying degrees of impeding the ascent of mantle-derived mafic magmas, which were produced by upwelling of asthenospheric mantle above the foundered Farallon slab

  16. Vegetative communities, Davis and Lavender Canyons, Paradox Basin, Utah: ecosystem studies

    International Nuclear Information System (INIS)

    1983-04-01

    The major vegetative communities of Davis and Lavender canyons located in southeastern Utah are characterized. The report identifies potential secondary impacts and appropriate mitigation options. The Davis Canyon and Lavender Canyon Study Area contains nine major vegetative communities: galleta-shadscale, juniper-blackbrush, juniper-shadscale-ephedra, shadscale-ephedra, grayia-shadscale, juniper, drywash, greasewood, and riparian. The natural recovery times of these communities are exceedingly long. Natural reinvasion of various species would take from 15 to 100 years. No threatened or endangered plant species were identified in the study area. Davis and Lavender canyons have been subject to off-road vehicle activity and extensive grazing. The plant communities may be subject to additional impacts as a result of increased human activity and off-highway activities such as camping, hiking, and hunting could result in changes in cover, composition, and frequency of plant species. Mitigation options for potential impacts include shuttle-busing workers to the site from the highway and fencing site access roads to prevent vehicles from leaving the roads

  17. 77 FR 29982 - Federal Acquisition Regulation; Submission for OMB Review; Davis Bacon Act-Price Adjustment...

    Science.gov (United States)

    2012-05-21

    ...; Submission for OMB Review; Davis Bacon Act-Price Adjustment (Actual Method) AGENCY: Department of Defense... previously approved information collection requirement concerning the Davis-Bacon Act price adjustment... Bacon Act-Price Adjustment (Actual Method), by any of the following methods: Regulations.gov : http...

  18. 77 FR 16548 - Florida Petroleum Reprocessors Superfund Site; Davie, Broward County, FL; Notice of Settlements

    Science.gov (United States)

    2012-03-21

    ...-2012- 3766; CERCLA-04-2012-3765] Florida Petroleum Reprocessors Superfund Site; Davie, Broward County... costs concerning the Florida Petroleum Reprocessors Superfund Site located in Davie, Broward County.... Painter. Submit your comments by Site name Florida Petroleum Reprocessors by one of the following methods...

  19. Davis PV plant operation and maintenance manual

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-09-01

    This operation and maintenance manual contains the information necessary to run the Photovoltaics for Utility Scale Applications (PVUSA) test facility in Davis, California. References to more specific information available in drawings, data sheets, files, or vendor manuals are included. The PVUSA is a national cooperative research and demonstration program formed in 1987 to assess the potential of utility scale photovoltaic systems.

  20. 76 FR 76438 - Agency Information Collection Activities; Submission for OMB Review; Comment Request; Davis-Bacon...

    Science.gov (United States)

    2011-12-07

    ... request (ICR) titled, ``Davis-Bacon Certified Payroll,'' to the Office of Management and Budget (OMB) for... for OMB Review; Comment Request; Davis-Bacon Certified Payroll ACTION: Notice. SUMMARY: The Department... of Management and Budget, Room 10235, Washington, DC 20503, Telephone: (202) 395-6929/Fax: (202) 395...

  1. Staging Unincorporated Power: Richard Harding Davis and the Critique of Imperial News

    Directory of Open Access Journals (Sweden)

    Nirmal Trivedi

    2011-12-01

    Full Text Available This essay contextualizes the work of war correspondent Richard Harding Davis within an evolving “imperial news apparatus” that would culminate in his reporting of the Spanish-American War. Critics have conventionally framed Davis squarely within the imperial cause, associating him with his admirer Roosevelt and naval admiral Alfred T. Mahan. Contrary to readings of Davis as an apologist for US imperialism, Trivedi contends that Davis understood how US imperial power relied on an information apparatus to communicate to an increasingly media-conscious American public through culture, that is, via familiar narratives, symbols, and objects—what Trivedi calls “imperial news.” The essay follows Davis’s development from his fictional representation of the new war correspondent in “The Reporter Who Made Himself King” to his own war correspondence before and after the Spanish-American War as collected in the memoirs A Year from a Reporter’s Notebook (1897, Cuba in War Time (1897, and Notes of a War Correspondent (1912. Davis’s war correspondence and fictional work effectively stage US imperialism as “unincorporated power”: that is, as power reliant on a developing news-making apparatus that deploys particular discursive strategies to validate its political claims. This staging critiques strategies of US imperial sovereignty—specifically its “privatization of knowledge” and its promotion of the war correspondent as nothing more than a spectator and purveyor of massacres.

  2. Staging Unincorporated Power: Richard Harding Davis and the Critique of Imperial News

    Directory of Open Access Journals (Sweden)

    Nirmal Trivedi

    2011-12-01

    Full Text Available This essay contextualizes the work of war correspondent Richard Harding Davis within an evolving “imperial news apparatus” that would culminate in his reporting of the Spanish-American War. Critics have conventionally framed Davis squarely within the imperial cause, associating him with his admirer Roosevelt and naval admiral Alfred T. Mahan. Contrary to readings of Davis as an apologist for US imperialism, Trivedi contends that Davis understood how US imperial power relied on an information apparatus to communicate to an increasingly media-conscious American public through culture, that is, via familiar narratives, symbols, and objects—what Trivedi calls “imperial news.” The essay follows Davis’s development from his fictional representation of the new war correspondent in “The Reporter Who Made Himself King” to his own war correspondence before and after the Spanish-American War as collected in the memoirs A Year from a Reporter’s Notebook (1897, Cuba in War Time (1897, and Notes of a War Correspondent (1912. Davis’s war correspondence and fictional work effectively stage US imperialism as “unincorporated power”: that is, as power reliant on a developing news-making apparatus that deploys particular discursive strategies to validate its political claims. This staging critiques strategies of US imperial sovereignty—specifically its “privatization of knowledge” and its promotion of the war correspondent as nothing more than a spectator and purveyor of massacres.

  3. Comparison of implementation of selected TMI action plan requirements on operating plants designed by Babcock and Wilcox

    International Nuclear Information System (INIS)

    Thoma, J.O.

    1984-05-01

    This report provides the results of a study conducted by the US Nuclear Regulatory Commission staff to compare the degree to which eight Babcock and Wilcox (B and W) designed licensed nuclear power plants have complied with the requirements in NUREG-0737, Clarification of TMI Action Plan Requirements. The eight licensed operating plants examined are as follows: Arkansas Nuclear One Unit 1 (ANO-1), Crystal River Unit 3, Davis Besse, Oconee Units 1, 2, and 3, Rancho Seco, and Three Mile Island Unit 1 (TMI-1). The purpose of this audit was to establish the progress of the TMI-1 licensee, General Public Utilities (GPU) Nuclear Corporation, in completing the long-term requirements in NUREG-0737 relative to the other B and W licensees examined

  4. 75 FR 76498 - Firstenergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Environmental...

    Science.gov (United States)

    2010-12-08

    ... will be no change to radioactive effluents that effect radiation exposures to plant workers and members... resources. There would be no impact to socioeconomic resources. Therefore, no changes to or different types...

  5. Training courses at VR-1 reactor

    International Nuclear Information System (INIS)

    Sklenka, L.; Kropik, M.

    2006-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilization - i.e. extensive educational program. The educational program is intended for the training of university students and selected nuclear power plant personnel. The training courses provide them experience in reactor and neutron physics, dosimetry, nuclear safety and operation of nuclear facilities. At present, the training course participants can go through more than 20 standard experimental exercises; particular exercises for special training can be prepared. Approximately 200 university students become familiar with the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. The VR-1 reactor takes also part in Eugene Wigner Course on Reactor Physics Experiments in the framework of European Nuclear Educational Network (ENEN) association. Recently, training courses for Bulgarian research reactor specialists supported by IAEA were carried out. An attractive program including demonstration of reactor operation is prepared also for high school students. Every year, more than 1500 high school students come to visit the reactor, as do many foreigner visitors. (author)

  6. Current status of the Thai Research Reactor (TRR-1/M1)

    International Nuclear Information System (INIS)

    Chueinta, Siripone; Julanan, Mongkol; Charncanchee, Decharchai

    2006-01-01

    The first Thai Research Reactor, TRR-1 went critical on 27 October 1962 at the maximum power of 1 MW. It was located at Office of Atoms for Peace (OAP) in Bangkok. Since then, TRR-1 was continuously operated and eventually shut down in 1975. Plate type, high-enriched uranium (HEU) and U 3 O 8 A1 cladding were used as the reactor fuel. Light water was used as moderator and coolant as well. In 1975, because of the problem from fuel supplier and also to supporting the Treaty of Non Proliferation of Nuclear Weapon or NPT, TRR-1 was shut down for modification. The reactor core and control system were disassembled and replaced by TRIGA Mark III. A new core was a hexagonal core shape designed by General Atomics (GA). Afterwards, TRR-1 was officially renamed to the Thai Research Reactor-1/Modification 1 (TRR-1/M1). TRR-1/M1 is a multipurpose swimming pool type reactor with nominal power of 2 MW. The TRR-1/M1 uses uranium enriched at 20% in U-235 (LEU) and ZrH alloy as fuel. Light water is also used as coolant and moderator. At present, the reactor is operating with core No.14. The reactor has been serving for various kinds of utilization namely, radioisotope production, neutron activation analysis, beam experiments and reactor physics experiments. (author)

  7. Potential noise impact from proposed operations at the Davis Canyon, Utah site: Evaluation of atmospheric acoustic refractive index profiles: Task 1, Final report

    International Nuclear Information System (INIS)

    Thomson, D.W.

    1986-01-01

    This study was motivated by the need to assess whether or not there would be significant noise impact from a proposed industrial operation to be sited in Davis Canyon, Utah. Completion of the study required improving several aspects of our fundamental understanding of atmospheric sound propagation and analysis of a diverse set of meteorological measurements which pertained specifically to the Davis Canyon location. The above two ''generic'' and ''specific'' objectives were sufficiently different that this final report has been divided into two parts. The first, generic, portion was prepared because neither existing noise standards nor standard field measurement techniques adequately recognize the importance of normal atmospheric boundary layer structure and processes on the magnitude and variations of noise propagated out-of-doors. The second, specific, part of the report summarizes a variety of acoustically-oriented analyses of meteorological measurements made near Davis Canyon. The results in both parts of the report are based on sophisticated atmospheric analysis, boundary layer and propagation models. The presentation of time dependent ''maps'' of predicted sound pressure levels (also as a function of frequency and source-surrounding topography) represents a significant advance in the state-of-the-art of environmental noise analysis and prediction

  8. 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)

  9. UC Davis Fuel Cell, Hydrogen, and Hybrid Vehicle (FCH2V) GATE Center of Excellence

    Energy Technology Data Exchange (ETDEWEB)

    Erickson, Paul

    2012-05-31

    This is the final report of the UC Davis Fuel Cell, Hydrogen, and Hybrid Vehicle (FCH2V) GATE Center of Excellence which spanned from 2005-2012. The U.S. Department of Energy (DOE) established the Graduate Automotive Technology Education (GATE) Program, to provide a new generation of engineers and scientists with knowledge and skills to create advanced automotive technologies. The UC Davis Fuel Cell, Hydrogen, and Hybrid Vehicle (FCH2V) GATE Center of Excellence established in 2005 is focused on research, education, industrial collaboration and outreach within automotive technology. UC Davis has had two independent GATE centers with separate well-defined objectives and research programs from 1998. The Fuel Cell Center, administered by ITS-Davis, has focused on fuel cell technology. The Hybrid-Electric Vehicle Design Center (HEV Center), administered by the Department of Mechanical and Aeronautical Engineering, has focused on the development of plug-in hybrid technology using internal combustion engines. The merger of these two centers in 2005 has broadened the scope of research and lead to higher visibility of the activity. UC Davis's existing GATE centers have become the campus's research focal points on fuel cells and hybrid-electric vehicles, and the home for graduate students who are studying advanced automotive technologies. The centers have been highly successful in attracting, training, and placing top-notch students into fuel cell and hybrid programs in both industry and government.

  10. Jefferson Davis and the Failure of Confederate Military Strategy, 1861-1865

    Science.gov (United States)

    2010-04-14

    seventeen. Here he first met some of the men who later held key roles in the Confederate Army: Albert Sidney Johnston, Leonidas Polk, and later Robert...General Lee’s magnificent Army ofNorthem Virginia. Throughout the war Jefferson Davis failed to assign to any theater and entirely subsidiary role ...fight. 24 0 ENDNOTES 1 A. H. McDannald, B.L., ed., The Modern Concise Encyclopedia, vol. III, (New York: Unicorn Press, 1941), 490-491. 2 Harry

  11. 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)

  12. Brownfields Davis Bacon for Cleanup Grants: Petroleum for Government Entities

    Science.gov (United States)

    Terms & conditions specify how Recipients will assist EPA in meeting its Davis Bacon responsibilities when DB applies to EPA awards of financial assistance under the Recovery Act or any other statute which makes DB applicable to EPA financial assistance.

  13. Safety operation of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    2001-01-01

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

  14. 10 CFR 455.112 - Davis-Bacon wage rate requirement.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 3 2010-01-01 2010-01-01 false Davis-Bacon wage rate requirement. 455.112 Section 455.112 Energy DEPARTMENT OF ENERGY ENERGY CONSERVATION GRANT PROGRAMS FOR SCHOOLS AND HOSPITALS AND BUILDINGS... acquisition and installation of more than $5,000, any construction contract or subcontract in excess of $2,000...

  15. Late-time structure of the Bunch-Davies de Sitter wavefunction

    Energy Technology Data Exchange (ETDEWEB)

    Anninos, Dionysios [Stanford Institute of Theoretical Physics, Stanford University, Stanford (United States); Anous, Tarek [Center for Theoretical Physics, Massachusetts Institute of Technology, Cambridge (United States); Freedman, Daniel Z. [Stanford Institute of Theoretical Physics, Stanford University, Stanford (United States); Center for Theoretical Physics, Massachusetts Institute of Technology, Cambridge (United States); Department of Mathematics, Massachusetts Institute of Technology, Cambridge (United States); Konstantinidis, George [Stanford Institute of Theoretical Physics, Stanford University, Stanford (United States)

    2015-11-30

    We examine the late time behavior of the Bunch-Davies wavefunction for interacting light fields in a de Sitter background. We use perturbative techniques developed in the framework of AdS/CFT, and analytically continue to compute tree and loop level contributions to the Bunch-Davies wavefunction. We consider self-interacting scalars of general mass, but focus especially on the massless and conformally coupled cases. We show that certain contributions grow logarithmically in conformal time both at tree and loop level. We also consider gauge fields and gravitons. The four-dimensional Fefferman-Graham expansion of classical asymptotically de Sitter solutions is used to show that the wavefunction contains no logarithmic growth in the pure graviton sector at tree level. Finally, assuming a holographic relation between the wavefunction and the partition function of a conformal field theory, we interpret the logarithmic growths in the language of conformal field theory.

  16. Anmeldelse: Whitney Davis A General Theory of Visual Culture

    DEFF Research Database (Denmark)

    Michelsen, Anders Ib

    2012-01-01

    Whitney Davis bog A General Theory of Visual Culture vil utvivlsomt blive opfattet som en provokation af mange deltagere i forskningsdebatterne om visuel kultur. At basere en »generel« teori om visuel kultur – dvs. en teori, som benytter sig af termer som »visualitet« – på et kerneargument de facto...

  17. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, Karel; Sklenka, Lubomir

    2003-01-01

    Full text: The training reactor VR-1 Vrabec ('Sparrow'), operated at the Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University in Prague, was started up on December 3, 1990. Particularly, it is designed and operated for training of students from Czech universities, preparing of experts for the Czech nuclear programme, as well as for certain research and development work, and for information programmes in the sphere of non-military nuclear energy use (public relation). The VR-1 training reactor is a pool-type light-water reactor based on enriched uranium with maximum thermal power 1kWth and short time period up to 5kW th . The moderator of neutrons is light demineralized water (H 2 O) that is also used as a reflector, a biological shielding, and a coolant. Heat is removed from the core with natural convection. The reactor core contains 14 to 18 fuel assemblies IRT-3M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The core is accommodated in a cylindrical stainless steel vessel - pool, which is filled with water. UR-70 control rods serve the reactor control and safe shutdown. Training of the VR-1 reactor provides students with experience in reactor and neutron physics, dosimetry, nuclear safety, and nuclear installation operation. Students from technical universities and from natural sciences universities come to the reactor for training. Approximately 200 university students are introduced to the reactor (lectures, experiments, experimental and diploma works, etc.) every year. About 12 different faculties from Czech universities use the reactor. International co-operation with European universities in Germany, Hungary, Austria, Slovakia, Holland and UK is frequent. Practical Course on Reactor Physics in Framework of European Nuclear Engineering Network has been newly introduced. Currently, students can try out more than 20 experimental exercises. Further training courses have been included

  18. Safety evaluation report related to Babcock and Wilcox Owners Group Plant Reassessment Program: [Final report

    International Nuclear Information System (INIS)

    1987-11-01

    After the accident of Three Mile Island, Unit 2, nuclear power plant owners made a number of improvements to their nuclear facilities. Despite these improvements, the US Nuclear Regulatory Commission (NRC) staff is concerned that the number and complexity of events at Babcock and Wilcox (B and W) nuclear plants have not decreased as expected. This concern was reinforced by the June 9, 1985 total-loss-of-feedwater event at Davis-Besse Nuclear Power Station and the December 26, 1985 overcooling transient at Rancho Seco Nuclear Generating Station. By letter dated January 24, 1986, the Executive Director for Operations (EDO) informed the Chairman of the B and W Owners Group (BWOG) that a number of recent events at B and W-designed reactors have led the NRC staff to conclude that the basic requirements for B and W reactors need to be reexamined. In its February 13, 1986 response to the EDO's letter, the BWOG committed to lead an effort to define concerns relative to reducing the frequency of reactor trips and the complexity of post-trip response in B and W plants. The BWOG submitted a description of the B and W program entitled ''Safety and Performance Improvement Program'' (BAW-1919) on May 15, 1986. Five revisions to BAW-1919 have also been submitted. The NRC staff has reviewed BAW-1919 and its revisions and presents its evaluation in this report. 2 figs., 34 tabs

  19. Brownfields Davis Bacon for Cleanup Grants: Hazardous Substances for Government Entities

    Science.gov (United States)

    The following terms and conditions specify how Recipients will assist EPA in meeting its Davis-Bacon (DB) responsibilities when DB applies to EPA awards of financial assistance under any statute which makes DB applicable to EPA financial assistance.

  20. Moving ring reactor 'Karin-1'

    International Nuclear Information System (INIS)

    1983-12-01

    The conceptual design of a moving ring reactor ''Karin-1'' has been carried out to advance fusion system design, to clarify the research and development problems, and to decide their priority. In order to attain these objectives, a D-T reactor with tritium breeding blanket is designed, a commercial reactor with net power output of 500 MWe is designed, the compatibility of plasma physics with fusion engineering is demonstrated, and some other guideline is indicated. A moving ring reactor is composed mainly of three parts. In the first formation section, a plasma ring is formed and heated up to ignition temperature. The plasma ring of compact torus is transported from the formation section through the next burning section to generate fusion power. Then the plasma ring moves into the last recovery section, and the energy and particles of the plasma ring are recovered. The outline of a moving ring reactor ''Karin-1'' is described. As a candidate material for the first wall, SiC was adopted to reduce the MHD effect and to minimize the interaction with neutrons and charged particles. The thin metal lining was applied to the SiC surface to solve the problem of the compatibility with lithium blanket. Plasma physics, the engineering aspect and the items of research and development are described. (Kako, I.)

  1. 42 CFR 137.379 - Do Davis-Bacon wage rates apply to construction projects performed by Self-Governance Tribes...

    Science.gov (United States)

    2010-10-01

    ... projects performed by Self-Governance Tribes using Federal funds? 137.379 Section 137.379 Public Health... HEALTH AND HUMAN SERVICES TRIBAL SELF-GOVERNANCE Construction Other § 137.379 Do Davis-Bacon wage rates apply to construction projects performed by Self-Governance Tribes using Federal funds? Davis-Bacon Act...

  2. Brownfields Davis Bacon for Cleanup Grants: Petroleum for Non-Profit Entities

    Science.gov (United States)

    Terms & conditions specify how Recipients will assist EPA in meeting its Davis Bacon responsibilities when DB applies to EPA awards of financial assistance under the Recovery Act or any other statute which makes DB applicable to EPA financial assistance.

  3. 75 FR 14635 - FirstEnergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Environmental...

    Science.gov (United States)

    2010-03-26

    ... be no change to radioactive effluents that effect radiation exposures to plant workers and members of... cultural resources. There would be no impact to socioeconomic resources. Therefore, no changes to or...

  4. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  5. Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

    International Nuclear Information System (INIS)

    Henry, A.F.

    1980-01-01

    Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

  6. The Davis-Bacon and Service Contract Acts: Laws Whose Time has Passed

    National Research Council Canada - National Science Library

    Pendolino, Timothy J

    1994-01-01

    In the Davis-Bacon and Service-Contract Acts, Congress attempted to protect the wages of workers in the construction and service industries by establishing a sort of minimum wage for Government contracts...

  7. Occupational analysis for the Angra-1 reactor

    International Nuclear Information System (INIS)

    Moraes, A.

    1991-01-01

    Due to several modifications which were imposed to its time schedule during construction, the Angra-1 reactor did not enter to the grid in 1982 as it was initially foreseen. These modifications occurred due to an unforeseen scenario that was verified in steam generators (serie D-3, Westinghouse) of power stations with similar configurations which had been installed in other countries such as Ringhals-3 (Sweden), Almaraz-1 (Spain) and McGuine-1 (USA). So, among the main events that occurred in the Angra-1 reactor, which were of interest from the point of view of radiation protection, it could be pointed out the personnel monitoring, and the occupational exposure measurements at different reactor power, during the reactor fueling and during modification and tests performed at the steam generators and at ducts of the primary coolant circuit. (author)

  8. A Critical Study of C.F.Davis's Views on Revelatory Religious Experiences

    Directory of Open Access Journals (Sweden)

    Ali Shirvani

    2010-12-01

    Full Text Available Revelatory experiences which are regarded as a major type of religious experiences comprise what their subjects may call sudden convictions, inspiration, revelation, enlightenment, the mystical vision and flashes of insight. In Davis's point of view, these experiences have distinctive features: (i They are usually sudden and of short duration (ii the alleged new knowledge seems to the subject to have been acquired not through reasoning or sense perception (iii the alleged new knowledge usually seems to the subject to have been ' poured into ' or ' showered upon ' him her by an external agency (iv the revelations carry with them utter convicition (v the gained insights are often claimed to be impossible to put into words.  This paper is to present how Davis describes religious experiences of this category from a Christian philosophical approach to religion. It also studies her point of views from a critical Islamic mystical vision. Through this critical and comparative study, it would be revealed that what Davis claimed to be known as revelatory religious experiences has close relationship with what is called "Kashfe Ma'lanavi" (spiritual intuition in Islamic mysticism. These are examined closely in Muslims' mystical sources. Distinction between Prophet's revelation and other instances of revelatory religious experiences and exploring their main differences were of close attention for Muslim mystics (Orafa.

  9. Environmental Survey preliminary report, Laboratory for Energy-Related Health Research, Davis, California

    International Nuclear Information System (INIS)

    1988-03-01

    This report presents the preliminary findings from the first phase of the Survey of the United States Department of Energy (DOE) Laboratory for Energy-Related Health Research (LEHR) at the University of California, Davis (UC Davis), conducted November 16 through 20, 1987. The Survey is being conducted by an interdisciplinary team of environmental specialists, led and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. Individual team components are being supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental risk associated with the LEHR. The Survey covers all environmental media and all areas of environmental regulation, and is being performed in accordance with the DOE Environmental Survey Manual. This phase of the Survey involves the review of existing site environmental data, observations of the operations at the LEHR and interviews with site personnel. The Survey team developed a Sampling and Analysis Plan to assist in further assessing certain of the environmental problems identified during its on-site activities. The Sampling and Analysis Plan will be executed by a DOE National Laboratory or a support contractor. When completed, the results will be incorporated into the Environmental Survey Interim Report for the LEHR at UC Davis. The Interim Report will reflect the final determinations of the LEHR Survey. 75 refs., 26 figs., 23 tabs

  10. Reactor operations at SAFARI-1

    International Nuclear Information System (INIS)

    Vlok, J.W.H.

    2003-01-01

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

  11. Light-water reactor research and development

    International Nuclear Information System (INIS)

    1985-05-01

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

  12. When did legong start? A reply to Stephen Davies

    Directory of Open Access Journals (Sweden)

    Adrian Vickers

    2009-05-01

    Full Text Available Stephen Davies has recently opened up new ways of looking at the history of Bali’s premier dance form, legong. He has argued that legong started in the late nineteenth century, more specifically after 1887, probably in 1889, and that it is primarily derived from a form which Balinese presently call andir. Davies’ article involves a substantial reconsideration of the canonical nature of certain dance forms in Bali. The evidence Davies used is largely oral history. What is missing from Davies’ account is evidence from closer to the time period, evidence that can allow us to fix the date of the origins of legong more closely, and also to understand precisely what its performative and musical associations and origins might be. This evidence is present in Balinese and Dutch-language sources, and while there are limitations to these sources, they certainly modify Davies’ thesis.

  13. 33 CFR 100.1102 - Marine Events on the Colorado River, between Davis Dam (Bullhead City, Arizona) and Headgate Dam...

    Science.gov (United States)

    2010-07-01

    ... River, between Davis Dam (Bullhead City, Arizona) and Headgate Dam (Parker, Arizona). 100.1102 Section... MARINE PARADES SAFETY OF LIFE ON NAVIGABLE WATERS § 100.1102 Marine Events on the Colorado River, between Davis Dam (Bullhead City, Arizona) and Headgate Dam (Parker, Arizona). (a) General. Sponsors are...

  14. Extensive utilization of training reactor VR-1

    International Nuclear Information System (INIS)

    Karel, Matejka; Lubomir, Sklenka

    2005-01-01

    This paper describes one of the main purposes of the VR-1 training reactor utilisation - i.e. extensive educational programme. The educational programme is intended for the training of university students (all technical universities in Czech Republic) and selected nuclear power plant personnel. At the present, students can go through more than 20 different experimental exercises. An attractive programme including demonstration of reactor operation is prepared also for high school students. Moreover, research and development works and information programmes proceed at the VR-1 reactor as well

  15. EOP and SAA Undergraduates Who Left UC Davis without a Degree.

    Science.gov (United States)

    Rasor, Marianne

    Undergraduate students enrolled in the Educational Opportunity Program (EOP) or the Student Affirmative Action (SAA) program at the University of California (UC), Davis, who withdrew before graduation were surveyed in 1981. Attention was directed to the respondents' educational experiences after leaving, their current employment, and their…

  16. Colorings of simplicial complexes and vector bundles over Davis-Januszkiewicz spaces

    NARCIS (Netherlands)

    Notbohm, D.R.A.W.

    2010-01-01

    We show that coloring properties of a simplicial complex K are reflected by splitting properties of a bundle over the associated Davis-Januszkiewicz space whose Chern classes are given by the elementary symmetric polynomials in the generators of the Stanley-Reisner algebra of K. © 2009 The

  17. Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

    2017-07-01

    Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

  18. Stability analysis of the Ghana Research Reactor-1 (GHARR-1)

    International Nuclear Information System (INIS)

    Della, R.; Alhassan, E.; Adoo, N.A.; Bansah, C.Y.; Nyarko, B.J.B.; Akaho, E.H.K.

    2013-01-01

    Highlights: • We developed a theoretical model to study the stability of the Ghana Research Reactor-1. • The neutronics transfer function was described by the point kinetics model for a single group of delayed neutrons. • The thermal hydraulics transfer function was based on the modified lumped parameter concept. • A computer code, RESA (REactor Stability Analysis) was developed. • Results show that the closed-loop transfer function was stable and well damped for variable operating power levels. - Abstract: A theoretical model has been developed to study the stability of the Ghana Research Reactor one (GHARR-1). The closed-loop transfer function of GHARR-1 was established based on the model, which involved the neutronics and the thermal hydraulics transfer functions. The reactor kinetics was described by the point kinetics model for a single group of delayed neutrons, whilst the thermal hydraulics transfer function was based on the modified lumped parameter concept. The inherent internal feedback effect due to the fuel and the coolant was represented by the fuel temperature coefficient and the moderator temperature coefficient respectively. A computer code, RESA (REactor Stability Analysis), entirely in Java was developed based on the model for systems analysis. Stability analysis of the open-loop transfer function of GHARR-1 based on the Nyquist criterion and Bode diagrams using RESA, has shown that the closed-loop transfer function was marginally stable for variable operating power levels. The relative stability margins of GHARR-1 were also identified

  19. Brownfields Davis Bacon for Cleanup Grants: Hazardous Substances for Non-Profit Entities

    Science.gov (United States)

    The following terms and conditions specify how Recipients will assist EPA in meeting its Davis Bacon (DB) responsibilities when DB applies to EPA awards of financial assistance under any other statute which makes DB applicable to EPA financial assistance.

  20. Tightness Entropic Uncertainty Relation in Quantum Markovian-Davies Environment

    Science.gov (United States)

    Zhang, Jun; Liu, Liang; Han, Yan

    2018-05-01

    In this paper, we investigate the tightness of entropic uncertainty relation in the absence (presence) of the quantum memory which the memory particle being weakly coupled to a decohering Davies-type Markovian environment. The results show that the tightness of the quantum uncertainty relation can be controlled by the energy relaxation time F, the dephasing time G and the rescaled temperature p, the perfect tightness can be arrived by dephasing and energy relaxation satisfying F = 2G and p = 1/2. In addition, the tightness of the memory-assisted entropic uncertainty relation and the entropic uncertainty relation can be influenced mainly by the purity. While in memory-assisted model, the purity and quantum correlation can also influence the tightness actively while the quantum entanglement can influence the tightness slightly.

  1. Population ecology of polar bears in Davis Strait, Canada and Greenland

    Science.gov (United States)

    Peacock, Elizabeth; Taylor, Mitchell K.; Laake, Jeffrey L.; Stirling, Ian

    2013-01-01

    Until recently, the sea ice habitat of polar bears was understood to be variable, but environmental variability was considered to be cyclic or random, rather than progressive. Harvested populations were believed to be at levels where density effects were considered not significant. However, because we now understand that polar bear demography can also be influenced by progressive change in the environment, and some populations have increased to greater densities than historically lower numbers, a broader suite of factors should be considered in demographic studies and management. We analyzed 35 years of capture and harvest data from the polar bear (Ursus maritimus) subpopulation in Davis Strait, including data from a new study (2005–2007), to quantify its current demography. We estimated the population size in 2007 to be 2,158 ± 180 (SE), a likely increase from the 1970s. We detected variation in survival, reproductive rates, and age-structure of polar bears from geographic sub-regions. Survival and reproduction of bears in southern Davis Strait was greater than in the north and tied to a concurrent dramatic increase in breeding harp seals (Pagophilus groenlandicus) in Labrador. The most supported survival models contained geographic and temporal variables. Harp seal abundance was significantly related to polar bear survival. Our estimates of declining harvest recovery rate, and increasing total survival, suggest that the rate of harvest declined over time. Low recruitment rates, average adult survival rates, and high population density, in an environment of high prey density, but deteriorating and variable ice conditions, currently characterize the Davis Strait polar bears. Low reproductive rates may reflect negative effects of greater densities or worsening ice conditions.

  2. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  3. From Stereotypes of Gender Difference to Stereotypes of Theory: A Response to Hayley Davis' Review of Deborah Tannen's "Gender and Discourse."

    Science.gov (United States)

    Yerian, Keli

    1997-01-01

    Responding to Hayley Davis' view of gender in discourse, asserts that she misinterprets Deborah Tannen as claiming that all men are well-intentioned and misunderstood, and that this misinterpretation is a theme appearing throughout Davis' review of Tannen's collection of essays, "Gender and Discourse". (29 references) (CK)

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  5. The Effects of the Davis Symbol Mastery System to Assist a Fourth Grader with Dyslexia in Spelling: A Case Report

    Science.gov (United States)

    Amsberry, Gianna; McLaughlin, T. F.; Derby, K. Mark; Waco, Teresa

    2012-01-01

    The purpose of this study was to determine the effectiveness of using the Davis Symbol Mastery Procedure for Words (Davis, 1994) for improving spelling skills. The participant was a fourth-grade male diagnosed with a significant learning disability. The intervention consisted of having the participant write each word, its definition, the word in a…

  6. Separation of magnetic affinity biopolymer adsorbents in a Davis tube magnetic separator

    Czech Academy of Sciences Publication Activity Database

    Šafařík, Ivo; Mucha, Pavel; Pechoč, Jiří; Stoklasa, Jaroslav; Šafaříková, Miroslava

    2001-01-01

    Roč. 23, - (2001), s. 851-855 ISSN 0141-5492 R&D Projects: GA ČR GA203/98/1145 Institutional research plan: CEZ:AV0Z6087904 Keywords : Davis tube * magnetic adsorbents * magnetic separation Subject RIV: EI - Biotechnology ; Bionics Impact factor: 0.915, year: 2001

  7. Burnup measurements at the RECH-1 research reactor

    International Nuclear Information System (INIS)

    Henriquez, C.; Navarro, G.; Pereda, C.; Torres, H.; Pena, L.; Klein, J.; Calderon, D.; Kestelman, A.J.

    2002-01-01

    The Chilean Nuclear Energy Commission has decided to produce LEU fuel elements for the RECH-1 research reactor. During December 1998, the Fuel Fabrication Plant delivered the first four fuel elements, called leaders, to the RECH-1 reactor. The set was introduced into the reactor's core, following the normal routine, but performing a special follow-up on their behavior inside and outside the core. In order to measure the burn-up of the leader fuel elements, it was decided to develop a burn-up measurements system to be installed into the RECH-1 reactor pool, and to decline the use of a similar system, which operates in a hot cell. The main reason to build this facility was to have the capability to measure the burn-up of fuel elements without waiting for long decay period. This paper gives a brief description of the facility to measure the burn-up of spent fuel elements installed into the reactor pool, showing the preliminary obtained spectra and briefly discussing them. (author)

  8. [Keratosis palmoplantaris maculosa seu papulosa (Davies-Colley) simulating multiple cornua cutanea].

    Science.gov (United States)

    Schreiber, D; Stücker, M; Hoffmann, K; Bacharach-Buhles, M; Altmeyer, P

    1997-08-01

    Patient with extensive keratosis palmoplantaris maculosa seu papulosa (Davies-Colley) presented with multiple cutaneous horns. The clinical picture, the histology, the electro microscopic examination, the negative tumor screening and the viral classification in the tissue allowed the differentiation from other palmoplantar keratoses. The patient was treated successfully using a combination of acitretin with physical and chemical measures.

  9. The effect of ileal interposition surgery on enteroendocrine cell numbers in the UC Davis type 2 diabetes mellitus rat

    DEFF Research Database (Denmark)

    Hansen, Carl Frederik; Vassiliadis, Efstathios; Vrang, Niels

    2014-01-01

    To investigate the short-term effect of ileal interposition (IT) surgery on gut morphology and enteroendocrine cell numbers in the pre-diabetic UC Davis type 2 diabetes mellitus (UCD-T2DM) rat.......To investigate the short-term effect of ileal interposition (IT) surgery on gut morphology and enteroendocrine cell numbers in the pre-diabetic UC Davis type 2 diabetes mellitus (UCD-T2DM) rat....

  10. Entangled de Sitter from stringy axionic Bell pair I. An analysis using Bunch-Davies vacuum

    International Nuclear Information System (INIS)

    Choudhury, Sayantan; Panda, Sudhakar

    2018-01-01

    In this work, we study the quantum entanglement and compute entanglement entropy in de Sitter space for a bipartite quantum field theory driven by an axion originating from Type IIB string compactification on a Calabi-Yau three fold (CY 3 ) and in the presence of an NS5 brane. For this computation, we consider a spherical surface S 2 , which divides the spatial slice of de Sitter (dS 4 ) into exterior and interior sub-regions. We also consider the initial choice of vacuum to be Bunch-Davies state. First we derive the solution of the wave function of the axion in a hyperbolic open chart by constructing a suitable basis for Bunch-Davies vacuum state using Bogoliubov transformation. We then derive the expression for density matrix by tracing over the exterior region. This allows us to compute the entanglement entropy and Renyi entropy in 3 + 1 dimension. Furthermore, we quantify the UV-finite contribution of the entanglement entropy which contain the physics of long range quantum correlations of our expanding universe. Finally, our analysis complements the necessary condition for generating non-vanishing entanglement entropy in primordial cosmology due to the axion. (orig.)

  11. Entangled de Sitter from stringy axionic Bell pair I. An analysis using Bunch-Davies vacuum

    Energy Technology Data Exchange (ETDEWEB)

    Choudhury, Sayantan [Inter-University Centre for Astronomy and Astrophysics, Pune (India); Tata Institute of Fundamental Research, Department of Theoretical Physics, Mumbai (India); Panda, Sudhakar [Institute of Physics, Bhubaneswar, Odisha (India); National Institute of Science Education and Research, Bhubaneswar, Odisha (India); Homi Bhabha National Institute, Mumbai (India)

    2018-01-15

    In this work, we study the quantum entanglement and compute entanglement entropy in de Sitter space for a bipartite quantum field theory driven by an axion originating from Type IIB string compactification on a Calabi-Yau three fold (CY{sup 3}) and in the presence of an NS5 brane. For this computation, we consider a spherical surface S{sup 2}, which divides the spatial slice of de Sitter (dS{sub 4}) into exterior and interior sub-regions. We also consider the initial choice of vacuum to be Bunch-Davies state. First we derive the solution of the wave function of the axion in a hyperbolic open chart by constructing a suitable basis for Bunch-Davies vacuum state using Bogoliubov transformation. We then derive the expression for density matrix by tracing over the exterior region. This allows us to compute the entanglement entropy and Renyi entropy in 3 + 1 dimension. Furthermore, we quantify the UV-finite contribution of the entanglement entropy which contain the physics of long range quantum correlations of our expanding universe. Finally, our analysis complements the necessary condition for generating non-vanishing entanglement entropy in primordial cosmology due to the axion. (orig.)

  12. Training and research on the nuclear reactor VR-1

    International Nuclear Information System (INIS)

    Matejka, K.

    1998-01-01

    The VR-1 training reactor is a light water reactor of the pool type using enriched uranium as the fuel. The moderator is demineralized light water, which also serves as the neutron reflector, biological shielding, and coolant. Heat evolved during the fission process is removed by natural convection. The reactor is used in the education of students in the field of reactor and neutron physics, dosimetry, nuclear safety, and instrumentation and control systems for nuclear facilities. Although primarily intended for students in various branches of technology (power engineering, nuclear engineering, physical engineering), this specialized facility is also used by students of faculties educating future natural scientists and teachers. Typical tasks trained at the VR-1 reactor include: measurement of delayed neutrons; examination of the effect of various materials on the reactivity of the reactor; measurement of the neutron flux density by various procedures; measurement of reactivity by various procedures; calibration of reactor control rods by various procedures; approaching the critical state; investigation of nuclear reactor dynamics; start-up, control and operation of a nuclear reactor; and investigation of the effect of a simulated nucleate boil on reactivity. In addition to the education of university-level students, training courses are also organized for specialists in the Czech nuclear programme

  13. Results of the Fall 2007 UC Davis Campus Travel Assessment

    OpenAIRE

    Congleton, Christopher

    2009-01-01

    Our collective transportation choices have far-reaching effects both locally and globally, from traffic congestion to global warming. While the concerted actions of many travelers working together could make significant inroads into solving these problems, a single traveler working alone could not. This report presents a snapshot of campus travel at the outset of the 2007-2008 academic year, measures campus mode split and average vehicle ridership, collects UC Davis travelers' opinions about ...

  14. The $100,000 Kiss: What Constitutes Peer Sexual Harassment for Schoolchildren under the "Davis v. Monroe County Board of Education" Holding?

    Science.gov (United States)

    Routh, Joanna L.

    1999-01-01

    Now that the Supreme Court in "Davis" has determined that schools can be sued for what one child does to another, schools will have a hard time avoiding frivolous lawsuits. The difficulty of analyzing the "Davis" decision lies in drawing a line between teasing and harassment. The conduct of certain six- and seven- year-olds…

  15. Thermal and hydraulic characteristics of the JEN-1 Reactor; Caracteristicas hidraulicas y termicas del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Otra Otra, F; Leira Rey, G

    1971-07-01

    In this report an analysis is made of the thermal and hydraulic performances of the JEN-1 reactor operating steadily at 3 Mw of thermal power. The analysis is made separately for the core, main heat exchanger and cooling tower. A portion of the report is devoted to predict the performances of these three main components when and if the reactor was going to operate at a power higher than the maximum 3 Mw attainable today. Finally an study is made of the unsteady operation of the reactor, focusing the attention towards the pumping characteristics and the temperatures obtained in the fuel elements. Reference is made to several digital calculation programmes that nave been developed for such purpose. (Author) 21 refs.

  16. IEA-R1 reactor - Spent fuel management

    International Nuclear Information System (INIS)

    Mattos, J.R.L. De

    1996-01-01

    Brazil currently has one Swimming Pool Research Reactor (IEA-R1) at the Instituto de Pesquisas Energeticas e Nucleares - Sao Paulo. The spent fuel produced is stored both at the Reactor Pool Storage Compartment and at the Dry Well System. The present situation and future plans for spent fuel storage are described. (author). 3 refs, 2 figs, 2 tabs

  17. Organic compounds in radiation fogs in Davis (California)

    Science.gov (United States)

    Herckes, Pierre; Hannigan, Michael P.; Trenary, Laurie; Lee, Taehyoung; Collett, Jeffrey L.

    New stainless steel active fogwater collectors were designed and used in Davis (CA, USA) to collect fogwater for the speciation of organic matter. Organic compounds in fog samples were extracted by liquid-liquid extraction and analyzed by gas chromatography coupled to mass spectrometry. Numerous organic compounds, including various alkanes, polycyclic aromatic hydrocarbons (PAH) and alkanoic acids, have been identified in the fogwater samples. Higher molecular weight (MW) compounds are preferentially associated with an insoluble phase inside the fog drops, whereas lower molecular weight and more polar compounds are found predominantly in the dissolved phase. Concentrations in the dissolved phase were sometimes much higher than estimated by the compounds' aqueous solubilities.

  18. Basic repository source term and data sheet report: Davis Canyon

    International Nuclear Information System (INIS)

    1988-01-01

    This report is one of series describing studies undertaken in support of the US Department of Energy Civilian Radioactive Waste Management (CRWM) Program. This study contains the derivation of values for environmental source terms and resources consumed for a CRWM repository. Estimates include heavy construction equipment; support equipment; shaft-sinking equipment; transportation equipment; and consumption of fuel, water electricity, and natural gas. Data are presented for construction and operation at an assumed site in Davis Canyon, Utah. 6 tabs

  19. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    Energy Technology Data Exchange (ETDEWEB)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

  20. TR-EDB: Test Reactor Embrittlement Data Base, Version 1

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

    1994-01-01

    The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

  1. Modernization and Refurbishment of the RECH-1 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Daie, J. [Nuclear Application Department, Chilean Nuclear Energy Commission (CCHEN), Santiago (Chile)

    2014-08-15

    The Chilean Nuclear Energy Commission (Comisión Chilena de Energía Nuclear, or CCHEN) has operated the RECH-1 research reactor since 1974. This reactor is located at La Reina Nuclear Centre in Santiago, Chile. It is a pool type reactor using LEU MTR fuel assemblies, light water as moderator and coolant, and beryllium as reflector. The reactor has been operated at the nominal power of 5 MW in a continuous shift of 20 hours per week, 48 weeks per year. The main utilizations of the RECH-1 reactor are radioisotope production and neutron activation analysis. Among the most relevant refurbishment and modernization campaigns undertaken at the reactor are: full core conversion to the use of LEU fuel, replacement of the cooling tower, improvement of the containment building by changing the doors and gates and by a better sealant for the penetrations, introduction of an additional source of water by connecting the raw water supply system to the emergency cooling system, improvement of the emergency ventilation system, introduction of a fire detection and alarm system for detection and mitigation to protect the I&C racks, introduction of a radioactive liquid release for those generated at the reactor, introduction of a delay tank degasification system and renewal of the environmental monitoring system. At present we are assessing the possibility of replacing the old analog electronics of control for new digital systems. Detailed descriptions of these diverse activities are presented in the paper. (author)

  2. Extensive utilisation of VR-1 reactor for nuclear education and training

    International Nuclear Information System (INIS)

    Rataj, J.

    2010-01-01

    The paper presents utilisation of the VR-1 reactor for nuclear education and training at national and international level. VR-1 reactor has been operating by the Czech Technical University since December 1990. The reactor is a pool-type light water reactor based on enriched uranium (19.7% 235 U) with maximum thermal power 1kW and for short time period up to 5kW. The moderator of neutrons is light water, which is also used as a reflector, a biological shielding and a coolant. Heat is removed from the core by natural convection. The pool disposition of the reactor facilitates access to the core, setting and removing of various experimental samples and detectors, easy and safe handling of fuel assemblies. The reactor core can contain from 17 to 21 fuel assemblies IRT-4M, depending on the geometric arrangement and kind of experiments to be performed in the reactor. The reactor is equipped with several experimental devices; e.g. horizontal, radial and tangential channels used to take out a neutron beam, reactivity oscillator for dynamics study and bubble boiling simulator. The reactor has been used very efficiently especially for education and training of university students and NPP's specialists for more than 18 years. The VR-1 reactor is utilised within various national and international activities such as Czech Nuclear Education Network (CENEN), European Nuclear Education Network and also Eastern European Research Reactor Initiative (EERRI). The reactor is well equipped for education and training not only by the experimental facility itself but also by incessant development of training methods and improvement of education experiments. The education experiments can be combined into training courses attended by students according to their study specialization and knowledge level. The training programme is aimed to the reactor and neutron physics, dosimetry, nuclear safety, and control of nuclear installations. Every year, approximately 250 university students undergo

  3. The search for possible time variations in Davis' measurements of the argon production rate in the solar neutrino experiment

    International Nuclear Information System (INIS)

    Haubold, H.J.; Gerth, E.

    1985-01-01

    With the gradual accumulation of experimental data in the solar neutrino experiment of Davis and collaborators (runs 18-74 for 1970-1982), the question, whether there are time variations of the solar neutrino flux, is of renewed interest. We discuss the mathematical-numerical methods applied to the statistical analysis of Davis' argon-37 production rate up till now known in the literature. These methods are characterized by the arbitrary arrangement of the Davis data in a time series. We perform a certain Fourier transformation for unequally-spaced time series of the measuring data of the argon-37 production rate, discuss the discovered periods and give significance criteria with respect to each period. We find that all periods discussed in the literature are contained in our series of periods. Pointing out the more mathematical character of the time series analysis we emphasize the predominant significance of the detected periods. (author)

  4. 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)

  5. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  6. Use of the VR-1 ''Vrabec'' training reactor

    International Nuclear Information System (INIS)

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

    1994-01-01

    An overview is presented of the extent and ways of using the VR-1 training reactor, which is operated by the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague. A list and the characteristics of 16 problems developed for teaching purposes is given, and the 14 faculties and 2 research institutes participating in the teaching activities are listed. The reactor is used in the education and training of nuclear scientists and engineers. The instrumentation, experimental, handling and operating tools, as well as documentation and texts relating to the reactor are described. The following examples of the teaching activities are included: a guided visit to the operating reactor site, reactor dynamics study and delayed neutron measurement, training course, and the basic criticality experiment. Nuclear safety aspects (hypothetical accidents, quality control and system qualification demonstration, safety culture) are stressed during the education. The reactor department is involved in international cooperation projects. (J.B.). 3 refs

  7. Chemical aspects of pellet-cladding interaction in light water reactor fuel elements

    International Nuclear Information System (INIS)

    Olander, D.R.

    1982-01-01

    In contrast to the extensive literature on the mechanical aspects of pellet-cladding interaction (PCI) in light water reactor fuel elements, the chemical features of this phenomenon are so poorly understood that there is still disagreement concerning the chemical agent responsible. Since the earliest work by Rosenbaum, Davies and Pon, laboratory and in-reactor experiments designed to elucidate the mechanism of PCI fuel rod failures have concentrated almost exclusively on iodine. The assumption that this is the reponsible chemical agent is contained in models of PCI which have been constructed for incorporation into fuel performance codes. The evidence implicating iodine is circumstantial, being based primarily upon the volatility and significant fission yield of this element and on the microstructural similarity of the failed Zircaloy specimens exposed to iodine in laboratory stress corrosion cracking (SCC) tests to cladding failures by PCI

  8. Electrical system regulations of the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Mello, Jose Roberto de; Madi Filho, Tufic

    2013-01-01

    The IEA-R1 reactor of the Nuclear and Energy Research Institute (IPEN-CNEN/SP), is a research reactor open pool type, designed and built by the U.S. firm Babcock and Wilcox, having, as coolant and moderator, deionized light water and beryllium and graphite, as reflectors. Until about 1988, the reactor safety systems received power from only one source of energy. As an example, it may be cited the control desk that was powered only by the vital electrical system 220V, which, in case the electricity fails, is powered by the generator group: no-break 220V. In the years 1989 and 1990, a reform of the electrical system upgrading to increase the reactor power and, also, to meet the technical standards of the ABNT (Associacao Brasileira de Normas Tecnicas) was carried out. This work has the objective of showing the relationship between the electric power system and the IEA-R1 reactor security. Also, it demonstrates that, should some electrical power interruption occur, during the reactor operation, this occurrence would not start an accident event. (author)

  9. The "dirty weather" diaries of Reverend Richard Davis: insights about early colonial-era meteorology and climate variability for northern New Zealand, 1839-1851

    Science.gov (United States)

    Lorrey, Andrew M.; Chappell, Petra R.

    2016-03-01

    Reverend Richard Davis (1790-1863) was a colonial-era missionary stationed in the Far North of New Zealand who was a key figure in the early efforts of the Church Mission Society. He kept meticulous meteorological records for the early settlements of Waimate North and Kaikohe, and his observations are preserved in a two-volume set in the Sir George Grey Special Collections in the Auckland Central Library. The Davis diary volumes are significant because they constitute some of the earliest land-based meteorological measurements that were continually chronicled for New Zealand. The diary measurements cover nine years within the 1839-1851 time span that are broken into two parts: 1839-1844 and 1848-1851. Davis' meteorological recordings include daily 9 a.m. and noon temperatures and midday pressure measurements. Qualitative comments in the diary note prevailing wind flow, wind strength, cloud cover, climate variability impacts, bio-indicators suggestive of drought, and notes on extreme weather events. "Dirty weather" comments scattered throughout the diary describe disturbed conditions with strong winds and driving rainfall. The Davis diary entries coincide with the end of the Little Ice Age (LIA) and they indicate southerly and westerly circulation influences and cooler winter temperatures were more frequent than today. A comparison of climate field reconstructions derived from the Davis diary data and tree-ring-based winter temperature reconstructions are supported by tropical coral palaeotemperature evidence. Davis' pressure measurements were corroborated using ship log data from vessels associated with iconic Antarctic exploration voyages that were anchored in the Bay of Islands, and suggest the pressure series he recorded are robust and can be used as "station data". The Reverend Davis meteorological data are expected to make a significant contribution to the Atmospheric Circulation Reconstructions across the Earth (ACRE) project, which feeds the major data

  10. Modernization of control instrumentation and security of reactor IAN - R1

    International Nuclear Information System (INIS)

    Gonzalez, J. M.

    1993-01-01

    The program to modernize IAN-R1 research reactor control and safety instrumentation has been carried out considering two main aspects: updating safety philosophy requirements and acquiring the newest reactor control instrumentation controlled by computer, following the present criteria internationally recognized, for safety and reliable reactor operations and the latest developments of nuclear electronic technology. The new IAN-R1 reactor instrumentation consist of two wide range neutron monitoring channels, commanded by microprocessor a data acquisition system and reactor control, (controlled by computers). The reactor control desk is providing through two displays; all safety and control signals to the reactor operators; furthermore some signals like reactor power, safety and period signals are also showed on digital bar graphics, which are hard wired directly from the neutron monitoring channels

  11. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

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

  12. Perspectives on reactor safety. Revision 1

    International Nuclear Information System (INIS)

    Haskin, F.E.; Hodge, S.A.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course

  13. Perspectives on reactor safety. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States); Hodge, S.A. [Oak Ridge National Lab., TN (United States). Engineering Technology Div.

    1997-11-01

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor safety concepts. The course consists of five modules: (1) the development of safety concepts; (2) severe accident perspectives; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  14. 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)

  15. Advances in Reactor Physics, Mathematics and Computation. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-01-01

    These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume one, are divided into 6 sessions bearing on: - session 1: Advances in computational methods including utilization of parallel processing and vectorization (7 conferences) - session 2: Fast, epithermal, reactor physics, calculation, versus measurements (9 conferences) - session 3: New fast and thermal reactor designs (9 conferences) - session 4: Thermal radiation and charged particles transport (7 conferences) - session 5: Super computers (7 conferences) - session 6: Thermal reactor design, validation and operating experience (8 conferences).

  16. Stationary low power reactor No. 1 (SL-1) accident site decontamination ampersand dismantlement project

    International Nuclear Information System (INIS)

    Perry, E.F.

    1995-01-01

    The Army Reactor Area (ARA) II was constructed in the late 1950s as a test site for the Stationary Low Power Reactor No. 1 (SL-1). The SL-1 was a prototype power and heat source developed for use at remote military bases using a direct cycle, boiling water, natural circulation reactor designed to operate at a thermal power of 3,000 kW. The ARA II compound encompassed 3 acres and was comprised of (a) the SL-1 Reactor Building, (b) eight support facilities, (c) 50,000-gallon raw water storage tank, (d) electrical substation, (e) aboveground 1,400-gallon heating oil tank, (f) underground 1,000-gallon hazardous waste storage tank, and (g) belowground power, sewer, and water systems. The reactor building was a cylindrical, aboveground facility, 39 ft in diameter and 48 ft high. The lower portion of the building contained the reactor pressure vessel surrounded by gravel shielding. Above the pressure vessel, in the center portion of the building, was a turbine generator and plant support equipment. The upper section of the building contained an air cooled condenser and its circulation fan. The major support facilities included a 2,500 ft 2 two story, cinder block administrative building; two 4,000 ft 2 single story, steel frame office buildings; a 850 ft 2 steel framed, metal sided PL condenser building, and a 550 ft 2 steel framed decontamination and laydown building

  17. Reactor utilization, Part 1

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1981-01-01

    The reactor operating plan for 1981 was subject to the needs of testing operation with the 80% enriched fuel and was fulfilled on the whole. This annex includes data about reactor operation, review of shorter interruptions due to demands of the experiments, data about safety shutdowns caused by power cuts. Period of operation at low power levels was used mostly for activation analyses, and the operation at higher power levels were used for testing and regular isotope production. Detailed data about samples activation are included as well as utilization of the reactor as neutron source and the operating plan for 1982 [sr

  18. Refurbishment of Pakistan research reactor (PARR-1) for stainless steel lining of the reactor pool

    International Nuclear Information System (INIS)

    Salahuddin, A.; Israr, M.; Hussain, M.

    2002-01-01

    Pakistan Research Reactor-1 (PARR-1) is a pool-type research reactor. Reactor aging has resulted in the increase of water seepage from the concrete walls of the reactor pool. To stop the seepage, it was decided to augment the existing pool walls with an inner lining of stainless steel. This could be achieved only if the pool walls could be accessed unhindered and without excessive radiation doses. For this purpose a partial decommissioning was done by removing all active core components including standard/control fuel elements, reflector elements, beam tubes, thermal shield, core support structure, grid plate and the pool's ceramic tiles, etc. An overall decommissioning program was devised which included procedures specific to each item. This led to the development of a fuel transport cask for transportation, and an interim fuel storage bay for temporary storage of fuel elements (until final disposal). The safety of workers and the environment was ensured by the use of specially designed remote handling tools, appropriate shielding and pre-planned exposure reduction procedures based on the ALARA principle. During the implementation of this program, liquid and solid wastes generated were legally disposed of. It is felt that the experience gained during the refurbishment of PARR-1 to install the stainless steel liner will prove useful and better planning and execution for the future decommissioning of PARR-1, in particular, and for other research reactors like PARR-2 (27 kW MNSR), in general. Furthermore, due to the worldwide activities on decommissioning, especially those communicated through the IAEA CRP on 'Decommissioning Techniques for Research Reactors', the importance of early planning has been well recognized. This has made possible the implementation of some early steps like better record keeping, rehiring of trained manpower, and creation of interim and final waste storage. (author)

  19. Reseña de libro: “16 de junio de 1955, bombardeos y masacre: imágenes, memorias, silencios” de Juan Besse y María Graciela Rodriguez (comp.)

    OpenAIRE

    Illanes, Marina

    2017-01-01

    El libro de Juan Besse y María Graciela Rodríguez nos presenta una mirada novedosa sobre el bombardeo a la Plaza de Mayo. El hecho fue prácticamente silenciado por los medios de comunicación, la historiografía y los distintos gobiernos que se sucedieron, hasta la década del 2000, particularmente a partir del 2005. Desde allí, encontramos varias investigaciones periodísticas sobre el tema, como los libros de Daniel Cichero y de Gonzalo Chaves, así como el primer informe estatal, realizado por ...

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  1. New instrumentation for the IPR-R1 reactor of CDTN

    International Nuclear Information System (INIS)

    Carvalho, P.V.R. de.

    1992-01-01

    The Nuclear Engineering Institute reactor instrumentation area has developed systems and equipment for reactor operation and safety. In such way, the new I and C for IEN Argonauta reactor and the nuclear instrumentation for IPEN critical facility were built. This paper describes our real work, the new I and C systems for IPR-R1, a Triga type reactor, located at CDTN (Belo Horizonte - MG). (author)

  2. Interview with a quality leader--Karen Davis, executive director of The Commonwealth Fund. Interview by Lecia A. Albright.

    Science.gov (United States)

    Davis, Karen

    2009-01-01

    Karen Davis is president of The Commonwealth Fund, a national philanthropy engaged in independent research on health and social policy issues. Dr. Davis is a nationally recognized economist, with a distinguished career in public policy and research. Before joining the Fund, she served as chairman of the Department of Health Policy and Management at The Johns Hopkins School of Public Health, where she also held an appointment as professor of economics. She served as deputy assistant secretary for health policy in the Department of Health and Human Services from 1977 to 1980, and was the first woman to head a U.S. Public Health Service agency. Before her government career, Ms. Davis was a senior fellow at the Brookings Institution in Washington, DC; a visiting lecturer at Harvard University; and an assistant professor of economics at Rice University. A native of Oklahoma, she received her PhD in economics from Rice University, which recognized her achievements with a Distinguished Alumna Award in 1991. Ms. Davis is the recipient of the 2000 Baxter-Allegiance Foundation Prize for Health Services Research. In the spring of 2001, Ms. Davis received an honorary doctorate in human letters from John Hopkins University. In 2006, she was selected for the Academy Health Distinguished Investigator Award for significant and lasting contributions to the field of health services research in addition to the Picker Award for Excellence in the Advancement of Patient Centered Care. Ms. Davis has published a number of significant books, monographs, and articles on health and social policy issues, including the landmark books HealthCare Cost Containment, Medicare Policy, National Health Insurance: Benefits, Costs, and Consequences, and Health and the War on Poverty. She serves on the Board of Visitors of Columbia University, School of Nursing, and is on the Board of Directors of the Geisinger Health System. She was elected to the Institute of Medicine (IOM) in 1975; has served two

  3. Comparison of the Wii Balance Board and the BESS tool measuring postural stability in collegiate athletes.

    Science.gov (United States)

    Guzman, Jill; Aktan, Nadine

    2016-02-01

    Concussions are a major health concern for athletes given the potential for these injuries in a wide range of sport activities. The leading concern for clinicians is that athletes are at risk for devastating consequences if they are not evaluated properly and cleared too early to return to play or competition. The evaluation of postural stability has been identified as an important aspect to the comprehensive management of such injuries. Clinicians are in need of a portable tool they can use in various settings to aid in decision making and health care delivery for concussed athletes. The Nintendo Wii Balance Board (Nintendo of America Inc., Redmond, Washington) is a portable, cost-effective tool that has the potential to aid in the evaluation of postural stability in concussed individuals. The purpose of this study was to evaluate the Wii Balance Board as an objective, user-friendly, cost effective, valid alternative tool for the measurement of postural stability in college athletes. This study questioned whether the Wii Balance Board, when compared to the Balance Error Scoring System (BESS), is an objective tool that can be used as an acceptable measurement of postural stability in college athletes. Copyright © 2015 Elsevier Inc. All rights reserved.

  4. 78 FR 13339 - Florida Petroleum Reprocessors Site; Davie, Broward County, FL; Notice of Settlement

    Science.gov (United States)

    2013-02-27

    ... ENVIRONMENTAL PROTECTION AGENCY [FRL 9785-7; CERCLA-04-2013-3755] Florida Petroleum Reprocessors... settlement with 2238 NW. 86th Street Inc. concerning the Florida Petroleum Reprocessors Site located in Davie... by Site name Florida Petroleum Reprocesssors Site by one of the following methods: www.epa.gov...

  5. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  6. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    Veloso, Marcelo Antonio; Fortini, Maria Auxiliadora

    2002-01-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  7. New human machine interface for VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Sklenka, L.; Chab, V.

    2002-01-01

    The contribution describes a new human machine interface that was installed at the VR-1 training reactor. The human machine interface update was completed in the summer 2001. The human machine interface enables to operate the training reactor. The interface was designed with respect to functional, ergonomic and aesthetic requirements. The interface is based on a personal computer equipped with two displays. One display enables alphanumeric communication between a reactor operator and the control and safety system of the nuclear reactor. Messages appear from the control system, the operator can write commands and send them there. The second display is a graphical one. It is possible to represent there the status of the reactor, principle parameters (as power, period), control rods' positions, the course of the reactor power. Furthermore, it is possible to set parameters, to show the active core configuration, to perform reactivity calculations, etc. The software for the new human machine interface was produced in the InTouch developing environment of the WonderWare Company. It is possible to switch the language of the interface between Czech and English because of many foreign students and visitors at the reactor. The former operator's desk was completely removed and superseded with a new one. Besides of the computer and the two displays, there are control buttons, indicators and individual numerical displays of instrumentation there. Utilised components guarantee high quality of the new equipment. Microcomputer based communication units with proper software were developed to connect the contemporary control and safety system with the personal computer of the human machine interface and the individual displays. New human machine interface at the VR-1 training reactor improves the safety and comfort of the reactor utilisation, facilitates experiments and training, and provides better support of foreign visitors.(author)

  8. Mike Davis, guérilla dans les sciences sociales.

    Directory of Open Access Journals (Sweden)

    Yveline Lévy-Piarroux

    2009-03-01

    Full Text Available « La plaine est morne et morte ― et la ville la mange. » Emile Verhaeren, « La plaine » in Les villes tentaculaires Le titre anglais de l’ouvrage de Mike Davis et Daniel B. Monk est Evil Paradises: Dreamworlds of Neoliberalism (2007. Les traducteurs ont conservé l’oxymore de la première formulation. Les « mondes de rêve » sont devenus « les villes hallucinées », allusion croisée à deux recueils de poésie d’Émile Verhaeren, Les campagnes hallucinées (1893 ...

  9. 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)

  10. NRDA-processed CTD data from the HOS Davis in the Gulf of Mexico, Cruise 1 Leg 1, collected from 2010-08-13 to 2010-08-22, associated with the Deepwater Horizon Oil Spill event (NCEI Accession 0128072)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Conductivity Temperature and Depth (CTD) measurements were collected aboard the R/V HOS Davis, Cruise 01, to determine physical oceanographic parameters of the water...

  11. Department of Reactor Technology annual progress report 1 January -31 December 1977

    International Nuclear Information System (INIS)

    1978-04-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, reactor operation, structural reliability, system reliability, reactor physics, fuel management, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, experimental activation measurements and neutron radiography at the DR 1 reactor, underground storage of gas, solar heating and underground heat storage, wind power. (author)

  12. Environmental Assessment Construction of a New Hazardous Cargo Pad Davis-Monthan AFB

    Science.gov (United States)

    2002-11-07

    PERFORMING ORGANIZATION NAME(S) AND ADDRESS(ES) 355th Civil Engineer Squadron (CES/CEVA),710 Third Street,Davis-Monthan AFB,AZ,85707 8. PERFORMING...agency for certain projects. Details of the preparation of this EA are mandated by the Council of Enviromental Quality (CEQ) in the series of...Base, Tucson, Arizona." October 1996. James M. Montgomery, Consulting Engineers for US army Corps of Engineers , Omaha Dist., Apri11990

  13. 75 FR 43519 - Parker-Davis Project; Transmission Capacity for Renewable Energy Between Electrical District No...

    Science.gov (United States)

    2010-07-26

    ... DEPARTMENT OF ENERGY Western Area Power Administration Parker-Davis Project; Transmission Capacity for Renewable Energy Between Electrical District No. 5 Substation and the Palo Verde Hub AGENCY... Department of Energy (DOE), is requesting SOIs from entities that are interested in purchasing transmission...

  14. 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)

  15. Thai Research Reactor (TRR-1/M1) Neutron Beam Measurements

    International Nuclear Information System (INIS)

    Ratanatongchai, Wichian

    2009-07-01

    Full text: Neutron beam tube of neutron radiography facility at Thai Research Reactor (TRR-1/M1) Thailand Institute of Nuclear Technology (public organization) is a divergent beam. The rectangular open-end of the beam tube is 16 cm x 17 cm while the inner-end is closed to the reactor core. The neutron beam size was measured using 20 cm x 40 cm neutron imaging plate. The measurement at the position 100 cm from the end of the collimator has shown that the beam size was 18.2 cm x 19.0 cm. Gamma ray in neutron the beam was also measured by the identical position using industrial X ray film. The area of gamma ray was 27.8 cm x 31.1 cm with the highest intensity found to be along the neutron beam circumference

  16. STS-47 MS Davis trains at Payload Crew Training Complex at Marshall SFC

    Science.gov (United States)

    1992-01-01

    STS-47 Endeavour, Orbiter Vehicle (OV) 105, Mission Specialist (MS) N. Jan Davis, wearing the Autogenic Feedback Training System 2 suit and lightweight headset, reviews a Payload Systems Handbook in the Spacelab Japan (SLJ) mockup during training at the Payload Crew Training Complex at Marshall Space Flight Center (MSFC) in Huntsville, Alabama. View provided with alternate number 92P-137.

  17. Türi viin : aasta jagu pohmellita võidujoovastust / Cathryne Davis ; interv. Neeme Raud

    Index Scriptorium Estoniae

    Davis, Cathryne

    2003-01-01

    Ameerika Ühendriikides müüdavatest Eesti viinadest on kõige suurema edu saavutanud Türi vodka. New-Yorgi suhtekorraldusfirma Deussen Global Comunications'i esindaja Cathryne Davis näeb müügiedu põhjustena viina kõrget kvaliteeti, turundusalast koostööd rahvusvaheliselt tuntud Bacardi'ga, eliidile suunatud sponsorlust, ulatuslikku müügivõrku ja edukat reklaamikampaaniat

  18. New measuring and protection system at VR-1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Jurickova, M.

    2006-01-01

    The contribution describes the new measuring and protection system of the VR-1 training reactor. The measuring and protection system upgrade is an integral part of the reactor I and C upgrade. The new measuring and protection system of the VR-1 reactor consists of the operational power measuring and the independent power protection systems. Both systems measure the reactor power and power rate, initiate safety action if safety limits are exceeded and send data (power, power rate, status, etc.) to the reactor control system. The operational power measuring system is a full power range system that receives signal from a fission chamber. The signal is evaluated according to the reactor power either in the pulse or current mode. The current mode utilizes the DC current and Campbell techniques. The new independent power protection system operates in the two highest reactor power decades. It receives signals from a boron chamber and evaluates it in the pulse mode. Both systems are computer based. The operational power measuring and independent power protection systems are diverse - different types and location of chambers, completely different hardware, software algorithms for the power and power rate calculations, software development tools and teems for the software manufacturing. (author)

  19. 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)

  20. 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)

  1. Dutton, Davies, and Imaginative Virtual Worlds: The Current State of Evolutionary Aesthetics

    Directory of Open Access Journals (Sweden)

    Joseph Carroll

    2013-12-01

    Full Text Available This paper is a commentary comparing the evolutionary perspectives of Denis Dutton’s The Art Instinct (2009 and Stephen Davies’s The Artful Species (2012. Their topics thus necessarily overlap, but their books have different purposes and a different feel. Davies’s book is an academic exercise. He has no real arguments or claims of his own. Dutton wishes to demonstrate that evolutionary psychology can provide a satisfying naturalistic explanation of aesthetic experience. Neither Davies nor Dutton fully succeeds in his ambition. Davies extends his scepticism well beyond a sensible account of the state of current knowledge about human evolution, and Dutton fails to recognize underlying theoretical differences in his main sources of theoretical inspiration. The limitations in these two works do not define the boundaries of current knowledge in evolutionary aesthetics. The most advanced and adequate concept in the evolutionary humanities is the idea that humans evolved the capacity to create imaginative virtual worlds and use those worlds to guide human behaviour. Both books being considered in this essay approach the idea of imaginative virtual worlds and almost grasp it. Before taking up that topic, the paper shall discuss two subsidiary issues: Dutton’s effort to incorporate sexual selection, and Davies’s sceptical negations about all evolutionary knowledge.

  2. Determination of Uranium in Aqueous and Organic Medium From Product and Waste Processes by Potentiometric Titration Using Modified Davies Gray Method

    International Nuclear Information System (INIS)

    Putro, K.P; Suripto, A

    1998-01-01

    Determination of uranium in aqueous and organic solution generated from nuclear fuels production and liquid radioactive waste at Fuel Element Production Installation for Research Reactor, by modified Davies-Gray method using phosphoric acid as medium and vanadium as catalyst has been carried out. The performed at different concentration of phosphoric acid, vanadium and the effect of impurities, as Al, Fe, Si, Cl and F in sample are measurement. Determination of uranium in organic solvent are the sample volume, agitation time and the optimum concentration of uranium to measurement. It was observed that, the optimum conditions for uranium analysis were : 5 -400 mg uranium in 3.2 M phosphoric acid medium containing 220 mg/l vanadium as catalyst. The impurities of Al ≤ 40.5 μg/ml, Fe ≤ 67.6 μg/ml, Si ≤ 20.3 μg/ml, Cl ≤ 135.1 μg/ml and F 13.5 μg/ml have not effect, but the concentration of F ≥ 40.5 μg/ml have effect in analysis result. The uranium content detectable in organic medium has been found between 0.01 to 0.10 g/l and the reproducibility range between 0.09 to 0.15 as well as the sample volume should be in the range of 5 and 10 ml by the agitation time for 4 minute

  3. Thermohydraulic analysis for power increase of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Umbehaun, Pedro E.; Bastos, Jose L.F.

    1996-01-01

    In this work has been presented the reactor core thermohydraulic model of IEAR-1, aiming its power operation increase from 2MW to 5MW. The design criteria adopted have been established in Safety Series 35. Three configurations of reactor core were analysed: fuel elements 20, 25 and 30

  4. Decommissioning and decontrolling the R1-reactor

    International Nuclear Information System (INIS)

    Bergman, C.; Holmberg, B.T.

    1985-01-01

    Sweden's first nuclear reactor - the research reactor R1 - situated in bedrock under the Royal Technical Institute of Stockholm, has in the period 1981-1983 been subject to a complete decommissioning. The National Institute for Radiation Protection has followed the work in detail, and has after the completion of the decommissioning performed measurements of radioactivity on site. The report gives an account of the work the Institute has done in preparation for- and during decommissioning and specifically report on the measurements for classification of the local as free for non-nuclear use. (aa)

  5. Late-time structure of the Bunch-Davies FRW wavefunction

    Science.gov (United States)

    Konstantinidis, George; Mahajan, Raghu; Shaghoulian, Edgar

    2016-10-01

    In this short note we organize a perturbation theory for the Bunch-Davies wavefunction in flat, accelerating cosmologies. The calculational technique avoids the in-in formalism and instead uses an analytic continuation from Euclidean signature. We will consider both massless and conformally coupled self-interacting scalars. These calculations explicitly illustrate two facts. The first is that IR divergences get sharper as the acceleration slows. The second is that UV-divergent contact terms in the Euclidean computation can contribute to the absolute value of the wavefunction in Lorentzian signature. Here UV divergent refers to terms involving inverse powers of the radial cutoff in the Euclidean computation. In Lorentzian signature such terms encode physical time dependence of the wavefunction.

  6. Upgrading of the research reactors FRG-1 and FRG-2

    International Nuclear Information System (INIS)

    Krull, W.

    1981-01-01

    In 1972 for the research reactor FRG-2 we applied for a license to increase the power from 15 MW to 21 MW. During this procedure a public laying out of the safety report and an upgrading procedure for both research reactors - FRG-1 (5 MW) and FRG-2 - were required by the licensing authorities. After discussing the legal background for licensing procedures in the Federal Republic of Germany the upgrading for both research reactors is described. The present status and future licensing aspects for changes of our research reactors are discussed, too. (orig.) [de

  7. 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

  8. Department of Reactor Technology: annual progress report 1 January - 31 December 1976

    International Nuclear Information System (INIS)

    1977-06-01

    The work of the Department of Reactor Technology within the following fields is described: reactor engineering, structural reliability, system reliability, radiation fiels in nuclear power plants, reactor physics, fuel management, fission product decay analysis, steady-state thermo-hydraulics, reactor accident analysis for LOCA and ECC, containment analysis, experimental heat transfer, reactor core dynamics and power plant simulators, control rod ejection accident analysis, economic studies for power plants, experimental activation measurements and neutron radiography at the DR 1 reactor. (author)

  9. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  10. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  11. Department of Reactor Technology annual progress report 1 January - 31 December 1978

    International Nuclear Information System (INIS)

    1979-04-01

    The activities of the department of reactor technology at Risoe during 1978 are described. The work is presented in five chapters: Reactor Engineering, Reactor Physics and Dynamics, Heat Transfer and Hydraulics, The DR 1 Reactor, and Non-Nuclear Activities. A list of the staff and of publications is included. (author)

  12. 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.)

  13. 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)

  14. 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.)

  15. 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)

  16. A Bioeconomic model of ocean acidification in the Baffin Bay/ Davis Strait Shrimp Fishery

    DEFF Research Database (Denmark)

    Kaiser, Brooks; Ravn-Jonsen, Lars

    We examine the case of the shrimp fishery in Baffin Bay/Davis Straight for potential effects of Ocean Acidification (OA), including: 1. the overall productivity of the shrimp fishery, 2. the spatial spread of the shrimp fishery, 3. the quality of the shrimp brought to market, and hence price...... and indirect costs of OA by comparing optimal bio-economic use of the shrimp fishery without ecosystem productivity shifts due to OA and with shifts due to OA. The demand side includes product differentiation to account for price differentials from different quality levels. The supply side includes costs...... or indirectly as the energy requirements of reproduction and growth shift the characteristics of the shrimp throughout the lifecycle....

  17. Modification of the IAN-R1 reactor

    International Nuclear Information System (INIS)

    Jaime, J.; Ahumada, S.; Spin, R.A.

    1990-01-01

    The IAN-R1 reactor is the only nuclear reactor operating in Colombia; it is installed at the Institute of Nuclear Affairs (AIN) in Bogota, which is an official body coming under the Ministry of Mining and Energy. This reactor started operation in January 1965 with a rated power of 10 kW and was modified a year later to operate at 20 kW, which has been its rated power up to the present. Given its importance for the application of nuclear technology in Columbia for various purposes, principally in the areas of neutron activation analysis, determination of uranium content in minerals using the delayed neutron counting method, production of certain radioisotopes such as 198 Au and 82 Br for engineering applications, and production of radioactive material for teaching and research purposes, research has been in progress for some years into ways of increasing its power. The study on experimental requirements and on the demand for locally produced radioisotopes came to the conclusion that its power should be increased to 1000 kW, which would allow the facility to remain on the same site. The modification includes conversion of the core to low-enriched fuel, operation up to 1 MW, modification of the shielding, renovation of instrumentation and installation of a radioisotope processing plant. When the reactor is modified we will be able to produce other radioisotopes for applications in nuclear medicine, industry and engineering; at the same time, the safety of the facility will be optimized and the experimental facilities improved

  18. Tracking spatial distribution of human-derived wastewater from Davis Station, East Antarctica, using δ15N and δ13C stable isotopes

    International Nuclear Information System (INIS)

    Corbett, Patricia A.; King, Catherine K.; Mondon, Julie A.

    2015-01-01

    Highlights: • Elevated δ15N and δ13C observed in fish tissue up to 4 km from the Davis Station wastewater outfall. • δ15N decreased stepwise with concentrations decreasing with distance from the discharge point. • The trend observed for δ13C almost mirrored δ15N. • Current wastewater treatment practices are insufficient to avoid uptake of contaminants in fish. - Abstract: Stable isotope ratios, δ15N and δ13C were effectively used to determine the geographical dispersion of human derived sewage from Davis Station, East Antarctica, using Antarctic rock cod (Trematomus bernacchii). Fish within 0–4 km downstream of the outfall exhibited higher δ15N and δ13C values relative to reference sites. Nitrogen in particular showed a stepped decrease in δ15N with increasing distance from the discharge point by 1–2‰. Stable isotopes were better able to detect the extent of wastewater contamination than other techniques including faecal coliform and sterol measures. Uptake and assimilation of δ15N and δ13C up to 4 km from the outfall adds to growing evidence indicating the current level of wastewater treatment at Davis Station is not sufficient to avoid impact to the surrounding environment. Isotopic assimilation in T. bernacchii is a viable biomarker for investigation of initial sewage exposure and longer term monitoring in the future

  19. California Environmental Vulnerability Assessment (CEVA) Score, San Joaquin Valley CA, 2013, UC Davis Center for Regional Change

    Data.gov (United States)

    U.S. Environmental Protection Agency — This data set is based on a three year study by the UC Davis Center for Regional Change, in affiliation with the Environmental Justice Project of the John Muir...

  20. Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Kassar, E.

    1986-01-01

    The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

  1. On-line chemistry monitoring for the secondary side

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Babcock and Wilcox (B and W) has developed a computerized water chemistry data acquisition and management system for nuclear plant secondary coolant systems. The Integrated Water Chemistry Monitoring System (IWCMS) provides on-line monitoring of conditions and rapid trend analysis of sampled data. So far it has been installed at GPU Three Mile Island unit 1 and at Toledo Edison Davis-Besse. The IWCMS meets the following utility needs for monitoring power plant chemistry: control of chemistry conditions to minimize corrosion and extend component/system life; continuous analysis of data from on-line detectors and grab samples; expediting of transient recovery actions with trend, alarm and evaluation capability; provision for rapid sharing of useful operational chemistry information; concentration of attention on evaluation instead of data manipulation. The system is composed of three functional parts: data acquisition hardware; PC-based computer system and customised system software. (author)

  2. Titanium Alloys for Critical Ordnance Components. Producers Coordination Meeting on Titanium Materials for Davy Crockett and Other Weapon Systems Held at Watertown Arsenal, Watertown 72, Mass., 14 Apr 60 and Fabricators Coordination Meeting on Titanium Materials for Davy Crockett and Other Weapon Systems Held at Watervliet Arsenal, Watervliet, NY, 15 Apr 1960

    Science.gov (United States)

    1960-01-01

    Title Agenda and AV^stracts - Watertown Arsenal, Watnrtovm, Mass. llj April i960 - Producers Coordination Meeting, "Titanium Materials for Davy...u.nd J.i’.l~.Jhr!l’ Scicnti~’ic Comp::1.~1y1 Chlc::-.~~o, IJ~~nois. oth :r c~tUl’ll~nt and 1Jl’Occ1t.G.~cn !.Jn.y be uoed if :J..CCC:t.> td )le to

  3. It Is My Desire to Be Free: Annie Davis's Letter to Abraham Lincoln and Winslow Homer's Painting "A Visit from the Old Mistress"

    Science.gov (United States)

    Hussey, Michael; Eder, Elizabeth K.

    2010-01-01

    "Mr. President, It is my Desire to be free," wrote Annie Davis to Abraham Lincoln, 20 months after he issued the Emancipation Proclamation. The Emancipation Proclamation affected only those parts of the country that were in rebellion against the United States on the date it was issued, January 1, 1863. The slaveholding border states of…

  4. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  5. 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)

  6. 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).

  7. 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)

  8. 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)

  9. 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)

  10. Review of ‘The Governance of Rangelands’ edited by Pedro M Herrera, Jonathan Davies and Pablo Manzano Baena

    Directory of Open Access Journals (Sweden)

    Claire Bedelian

    2017-11-01

    Full Text Available Book details Herrera, PM; Davies, J and Manzano Baena P The Governance of Rangelands: Collective action for sustainable pastoralism Routledge, London and New York, 2014 298 pages, ISBN 9781138785144

  11. Comment on Lockwood and Davis, "On the longitudinal extent of magnetopause reconnection pulses"

    Directory of Open Access Journals (Sweden)

    W. J. Heikkila

    1999-02-01

    Full Text Available Lockwood and Davis (1996 present a concise description of magnetopause reconnection pulses, with the claimed support of three types of observations: (1 flux transfer events (FTE, (2 poleward-moving auroral forms on the dayside, and (3 steps in cusp ion dispersion characteristics. However, there are a number of errors and misconceptions in the paper that make their conclusions untenable. They do not properly take account of the fact that the relevant processes operate in the presence of a plasma. They fail to notice that the source of energy (a dynamo with E · J<0 must be close to the region of dissipation (the electrical load with E · J>0 in transient phenomena, since energy (or information cannot travel faster than the group velocity of waves in the medium (here the Alfvén velocity VA. In short, Lockwood and Davis use the wrong contour in their attempt to evaluate the electromotive force (emf. This criticism goes beyond their article: a dynamo is not included in the usual definition of reconnection, only the reconnection load. Without an explicit source of energy in the assumed model, the idea of magnetic reconnection is improperly posed. Recent research has carried out a superposed epoch analysis of conditions near the dayside magnetopause and has found the dynamo and the load, both within the magnetopause current sheet. Since the magnetopause current is from dawn to dusk, the sign of E · J reflects the sign of the electric field. The electric field reverses, within the magnetopause; this can be discovered by an application of Lenz's law using the concept of erosion of the magnetopause. The net result is plasma transfer across the magnetopause to feed the low latitude boundary layer, at least partly on closed field lines, and viscous interaction as the mechanism by which solar wind plasma couples to the magnetosphere.

  12. BWR NSSS design basis documentation

    International Nuclear Information System (INIS)

    Vij, R.S.; Bates, R.E.

    2004-01-01

    In 1985 an incident at Toledo Edison's Davis Besse plant caused the U.S. Nuclear Regulatory Commission (NRC) to re-evaluate the technical information that the utilities had readily available to support the design of their plants. The Design Basis programs, currently on going in most U.S. utilities, have been the nuclear industry's response to the needs identified by this re-evaluation. In order to understand the Design Basis programs which have been implemented by the U.S. nuclear utilities, it is necessary to understand the problem as it was perceived by the nuclear industry (the utilities, the original NSSS designers and the regulators) after the Davis-Besse incident, the subsequent programs undertaken by the industry under the leadership of INPO and NUMARC, the NRC's actions, and the overall evolution of the industry's vision in relation to this problem. This paper presents the history of the design basis efforts from the first recognition of the problem by the NRC after the Davis-Besse incident, describes the actions taken by the NRC, INPO, NUMARC, the U.S. utilities and the NSSS designers, and brings the problem statement up-to-date in relation to the vision presently held by the U.S. nuclear industry. It then presents a technical discussion to develop a detailed definition of design basis information to support the problem statement. The information originally supplied by the NSSS designers during the plant design and construction is discussed as well as its relationship to the previously defined design basis information. This section of the paper concludes by defining the additional information needed by nuclear utilities to satisfy the requirements developed from the problem statement. Having developed a definition of the additional information (i.e., information not originally supplied during design and construction) required to solve the design basis problem as it is presently perceived by the U.S. nuclear industry, the paper then discusses design basis

  13. Present and future activities of TRIGA RC-1 Reactor

    International Nuclear Information System (INIS)

    Festinesi, A.

    1986-01-01

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  14. Neutronic studies in the enrichment reduction of research reactor IEAR-1

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Fanaro, L.C.B.; Mai, L.A.; Ferreira, P.S.B.; Garone, J.G.M.

    1987-01-01

    In the present work the codes used by the Reactor Physics Division of IPEN-CNEN-SP in calculations for plate-type reactors are described analyzing research reactor IEAR-1. The IAEA model problem for a plate-type reactor 10 MW with high, medium and low enrichment is solved through different methodologies now in use at the RTF/IPEN-CNEN-SP (HAMMER and HAMMER-TECH-CITATION and LEO4-2DBP-UM) looking into the calculation capability for high to low enrichment conversion within the contract held with the IAEA (BRA-4661). Finally, present reactor configuration calculations are compared with experimental measurements with the aim to validate the calculation method. (Author)

  15. 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

  16. 75 FR 7628 - Davis-Besse Nuclear Power Station; Notice of Consideration of Issuance of Amendment to Facility...

    Science.gov (United States)

    2010-02-22

    ...; Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing The U.S. Nuclear Regulatory Commission... involves no significant hazards consideration. Under the Commission's regulations in Title 10 of the Code...

  17. Rotenone formulation fate in Lake Davis following the 2007 treatment.

    Science.gov (United States)

    Vasquez, Martice E; Rinderneck, Janna; Newman, Julie; McMillin, Stella; Finlayson, Brian; Mekebri, Abdou; Crane, David; Tjeerdema, Ronald S

    2012-05-01

    In September 2007, Lake Davis (near Portola, California) was treated by the California Department of Fish and Game with CFT Legumine, a rotenone formulation, to eradicate the invasive northern pike (Esox lucius). The objective of this report is to describe the fate of the five major formulation constituents-rotenone, rotenolone, methyl pyrrolidone (MP), diethylene glycol monethyl ether (DEGEE), and Fennedefo 99-in water, sediment, and brown bullhead catfish (Ameiurus nebulosus; a rotenone-resistant species) by determination of their half-lives (t(1/2)) and pseudo first-order dissipation rate constants (k). The respective t(1/2) values in water for rotenone, rotenolone, MP, DEGEE, and Fennedefo 99 were 5.6, 11.1, 4.6, 7.7, and 13.5 d; in sediments they were 31.1, 31.8, 10.0, not able to calculate, and 48.5 d; and in tissues were 6.1, 12.7, 3.7, 3.2, and 10.4 d, respectively. Components possessing low water solubility values (rotenone and rotenolone) persisted longer in sediments (not detectable after 157 d) and tissues (<212 d) compared with water, whereas the water-miscible components (MP and DEGEE) dissipated more quickly from all matrices, except for Fennedefo 99, which was the most persistent in water (83 d). None of the constituents was found to bioaccumulate in tissues as a result of treatment. In essence, the physicochemical properties of the chemical constituents effectively dictated their fate in the lake following treatment. Copyright © 2012 SETAC.

  18. Reliabitity study of the accumulator system for Angra-1 reactor

    International Nuclear Information System (INIS)

    Santos Maciel, C.C.R.

    1980-01-01

    The realibility of the Accumulator System of Angra 1 reactor is studied. The fault tree techniques is use for identification and evaluation of the probability of occurrence of the possible failure modes of the system. The study has as a guide the report WASH 1400 in which the analysis of the reliability of a Tipical PWR reactor of USA. Comparisons between results obtained for Accumulator System of Angra 1 and that published in the report WASH 1400 for the Accumulator System of the Typical Reactor are done. Critiques to the methodology used in the reportd WASH 1400 and an analysis of the sensitivity of the system in relation with its components are also done. (author) [pt

  19. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.

    2004-01-01

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  20. Humphrey Davy and the Safety Lamp: The Use of Metal Gauze as a Flame Barrier

    Science.gov (United States)

    Mills, Allan

    2015-01-01

    The "safety lamp" invented by Humphrey Davy in 1815 utilised the cooling effect of metal gauze to prevent the flame of a candle or oil lamp (essential for illumination in mines) from passing through such a screen. It is therefore rendered unable to ignite any potentially explosive mixture of air and methane in the atmosphere surrounding…

  1. 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.)

  2. Graphite stack corrosion of BUGEY-1 reactor (synthesis)

    International Nuclear Information System (INIS)

    Petit, A.; Brie, M.

    1996-01-01

    The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs

  3. Sandia reactor kinetics codes: SAK and PK1D

    International Nuclear Information System (INIS)

    Pickard, P.S.; Odom, J.P.

    1978-01-01

    The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time

  4. Evaluation of air quality and noise impact assessments, Davis Canyon

    International Nuclear Information System (INIS)

    1986-05-01

    In this report, several issues are identified regarding the air quality and noise assessments presented in the final salt repository environmental assessment (EA) prepared by the US Department of Energy for the Davis Canyon, Utah, site. Necessary revisions to the data and methods used to develop the EA impact assessment are described. Then, a comparative evaluation is presented in which estimated impacts based upon the revised data and methods are compared with the impacts published in the EA. The evaluation indicates that the conclusions of the EA air quality and noise impact sections would be unchanged. Consequently, the guideline findings presented in Chapter 6 of the EA are also unchanged by the revised analysis. 50 refs., 16 tabs

  5. Multi-image screening technique applied to a general orientation training program

    International Nuclear Information System (INIS)

    Hajek, B.K.; Campbell, T.O.; Evans, A.D.; Hickey, J.M.

    1979-01-01

    A general orientation and training program is a prerequisite for personnel to have unescorted access to various site locations at a nuclear power plant. A new general orientation and training program is being developed for the Toledo Edison Company to be used at the Davis-Besse Nuclear Power Station. The program is presented in a multi-image and stereo sound format that has the unique capability to present the magnitude and scale of the plant, to arouse and maintain the interest of the viewer, and to instill in him a feeling of importance and pride about his job. Satisfactory completion of the program by individuals is assessed and certified by a machine scored test that is administered as an integral part of the presentation

  6. Davis Canyon noise analysis: Revision 2

    International Nuclear Information System (INIS)

    1985-11-01

    A study was performed as part of the Civilian Radioactive Waste Management Program to quantify the level and effect of noise from the various major phases of development of the proposed potentially acceptable nuclear waste repository site at Davis Canyon, Utah. This report contains the results of a predictive noise level study for the site characterization, repository construction, and repository operational phases. Included herein are graphic representations of energy averaged sound levels, and of audibility levels representing impact zones expected during each phase. Sound levels from onsite and offsite activity including traffic on highways and railroad routes are presented in isopleth maps. A description of the Environmental Noise Prediction Model used for the study, the study basis and methodologies, and actual modeling data are provided. Noise and vibration levels from blasting are also predicted and evaluated. Protective noise criteria containing a margin of safety are used in relation to residences, schools, churches, noise-sensitive recreation areas, and noise-sensitive biological resources. Protective ground motion criteria for ruins and delicate rock formation in Canyonlands National Park and for human annoyance are used in the evaluation of blasting. The evaluations provide the basis for assessing the noise impacts from the related activities at the proposed repository. 45 refs., 21 figs., 15 tabs

  7. The Hugh Davies Collection: live electronic music and self-built electro-acoustic musical instruments, 1967–1975

    Directory of Open Access Journals (Sweden)

    Dr James Mooney

    2017-04-01

    Full Text Available The Hugh Davies Collection (HDC at the Science Museum in London comprises 42 items of electronic sound apparatus owned by English experimental musician Hugh Davies (1943–2005, including self-built electro-acoustic musical instruments and modified sound production and manipulation hardware. An early proponent of ‘live electronic music’ (performed live on stage rather than constructed on magnetic tape in a studio, Davies’s DIY approach shaped the development of experimental and improvised musics from the late 1960s onwards. However, his practice has not been widely reported in the literature, hence little information is readily available about the material artefacts that constituted and enabled it. This article provides the first account of the development of Davies’s practice in relation to the objects in the HDC: from the modified electronic sound apparatus used in his early live electronic compositions (among the first of their kind by a British composer; through the ‘instrumental turn’ represented by his first self-built instrument, Shozyg I (1968; to his mature practice, where self-built instruments like Springboard Mk. XI (1974 replaced electronic transformation as the primary means by which Davies explored new and novel sound-worlds. As well as advancing knowledge of Davies’s pioneering work in live electronics and instrument-building and enhancing understanding of the objects in the HDC, this article shows how object biographic and archival methodologies can be combined to provide insight into the ways in which objects (instruments, technologies and practices shape each other over time.

  8. Ageing problems and renovation programme of ET-RR-1 reactor

    International Nuclear Information System (INIS)

    Khattab, M.S.; Sultan, M.A.

    1995-01-01

    Based on Practical Experience gained from interfacing ageing systems in addition to operating new systems, current problems could be deduced whenever in-service inspection are carried out. This paper summarizes the in-service inspection made, and the proposed programme of rehabilitation of mechanical system in the ET-RR-1 research reactor at Inshass. Exchangeable experience in solving common problems in similar reactors play an important role in the effectiveness of such rehabilitation programme. The paper summarizes also the modernization of control, measuring and radiation monitoring system already carried out at the reactor. (orig.)

  9. Generating the flux map of Nigeria Research Reactor-1 for efficient ...

    African Journals Online (AJOL)

    One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

  10. Basic repository environmental assessment design basis, Davis Canyon site

    International Nuclear Information System (INIS)

    1984-01-01

    This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Davis Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1,000-year-life carbon-steel waste packages in a collected waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.49 billion. Costs include those for the collocated WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 50 refs., 24 figs., 20 tabs

  11. Reactor protection system. Revision 1

    International Nuclear Information System (INIS)

    Fairbrother, D.B.; Vincent, D.R.; Lesniak, L.M.

    1975-04-01

    The reactor protection system-II (RPS-II) designed for use on Babcock and Wilcox 145- and 205-fuel assembly pressurized water reactors is described. In this system, relays in the trip logic have been replaced by solid state devices. A calculating module for the low DNBR, pump status, and offset trip functions has replaced the overpower trip (based on flow and imbalance), the power/RC pump trip, and the variable low pressure trip. Included is a description of the changes from the present Oconee-type reactor protection system (RPS-I), a functional and hardware description of the calculating module, and a discussion of the qualification program conducted to ensure that the degree of protection provided by RPS-II is not less than that provided by previously licensed systems supplied by B and W. (U.S.)

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  13. Using aetnanova to formally prove that the Davis-Putnam satisfiability test is correct

    Directory of Open Access Journals (Sweden)

    Eugenio G. Omodeo

    2008-05-01

    Full Text Available This paper reports on using the ÆtnaNova/Referee proof-verification system to formalize issues regarding the satisfiability of CNF-formulae of propositional logic. We specify an “archetype” version of the Davis-Putnam-Logemann-Loveland algorithm through the THEORY of recursive functions based on a well-founded relation, and prove it to be correct.Within the same framework, and by resorting to the Zorn lemma, we develop a straightforward proof of the compactness theorem.

  14. A variation of the Davis-Smith method for in-flight determination of spacecraft magnetic fields.

    Science.gov (United States)

    Belcher, J. W.

    1973-01-01

    A variation of a procedure developed by Davis and Smith (1968) is presented for the in-flight determination of spacecraft magnetic fields. Both methods take statistical advantage of the observation that fluctuations in the interplanetary magnetic field over short periods of time are primarily changes in direction rather than in magnitude. During typical solar wind conditions between 0.8 and 1.0 AU, a statistical analysis of 2-3 days of continuous interplanetary field measurements yields an estimate of a constant spacecraft field with an uncertainty of plus or minus 0.25 gamma in the direction radial to the sun and plus or minus 15 gammas in the directions transverse to the radial. The method is also of use in estimating variable spacecraft fields with gradients of the order of 0.1 gamma/day and less and in other special circumstances.

  15. Thirtieth anniversary of reactor accident in A-1 Nuclear Power Plant Jaslovske Bohunice

    International Nuclear Information System (INIS)

    Kuruc, J.; Matel, L.

    2007-01-01

    The facts about reactor accidents in A-1 Nuclear Power Plant Jaslovske Bohunice, Slovakia are presented. There was the reactor KS150 (HWGCR) cooled with carbon dioxide and moderated with heavy water. A-1 NPP was commissioned on December 25, 1972. The first reactor accident happened on January 5, 1976 during fuel loading. This accident has not been evaluated according to the INES scale up to the present time. The second serious accident in A-1 NPP occurred on February 22, 1977 also during fuel loading. This INES level 4 of reactor accident resulted in damaged fuel integrity with extensive corrosion damage of fuel cladding and release of radioactivity into the plant area. The A-1 NPP was consecutively shut down and is being decommissioned in the present time. Both reactor accidents are described briefly. Some radioecological and radiobiological consequences of accidents and contamination of area of A-1 NPP as well as of Manivier Canal and Dudvah River as result of flooding during the decommissioning are presented (authors)

  16. Demolition of the FRJ-1 research reactor (MERLIN)

    International Nuclear Information System (INIS)

    Stahn, B.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2003-01-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [de

  17. Evaporation rate measurement in the pool of IEAR-1 reactor

    International Nuclear Information System (INIS)

    Torres, Walmir Maximo; Cegalla, Miriam A.; Baptista Filho, Benedito Dias

    2000-01-01

    The surface water evaporation in pool type reactors affects the ventilation system operation and the ambient conditions and dose rates in the operation room. This paper shows the results of evaporation rate experiment in the pool of IEA-R1 research reactor. The experiment is based on the demineralized water mass variation inside cylindrical metallic recipients during a time interval. Other parameters were measured, such as: barometric pressure, relative humidity, environmental temperature, water temperature inside the recipients and water temperature in the reactor pool. The pool level variation due to water contraction/expansion was calculated. (author)

  18. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  19. IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

    International Nuclear Information System (INIS)

    Frajndlich, Roberto

    2009-01-01

    The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)

  20. 77 FR 14504 - University of California, Davis, et al.; Notice of Decision on Applications for Duty-Free Entry...

    Science.gov (United States)

    2012-03-12

    ... Shields Avenue Davis, CA 95616. Instrument: Alexsys 1000 Calorimeter. Manufacturer: Setaram... calorimeters, made by other companies, are completely different in design and do not feature the large sample volume surrounded by a sensitive detector that is essential for solution calorimetry. Docket Number: 12...

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. Thermal hydraulic and safety analyses for Pakistan Research Reactor-1

    International Nuclear Information System (INIS)

    Bokhari, I.H.; Israr, M.; Pervez, S.

    1999-01-01

    Thermal hydraulic and safety analysis of Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel have been performed using computer code PARET. The present core comprises of 29 standard and 5 control fuel elements. Results of the thermal hydraulic analysis show that the core can be operated at a steady-state power level of 10 MW for a flow rate of 950 m 3 /h, with sufficient safety margins against ONB (onset of nucleate boiling) and DNB (departure from nucleate boiling). Safety analysis has been carried out for various modes of reactivity insertions. The events studied include: start-up accident; accidental drop of a fuel element in the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results indicate that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is therefore concluded that the reactor can be safely operated at 10 MW without compromising safety. (author)

  3. Spent fuel management - two alternatives at the FiR 1 reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S.E.J.

    2001-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The reactor with its subsystems has experienced a large renovation work in 1996-97. The main purpose of the upgrading was to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The BNCT work dominates the current utilization of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The Council of State (government) granted for the reactor a new operating license for twelve years starting from the beginning of the year 2000. There is however a special condition in the new license. We have to achieve a binding agreement between our Research Centre and the domestic Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel, if we want to continue the reactor operation beyond the year 2006. In addition to the choosing of one of the spent fuel management alternatives the future of the reactor will also depend strongly on the development of the BNCT irradiations. If the number of patients per year increases fast enough and the irradiations of the patients will be economically justified, the operation of the reactor will continue independently of the closing of the USDOE alternative in 2006. Otherwise, if the number of patients will be low, the funding of the reactor will be probably stopped and the reactor will be shut down. (author)

  4. Measurement of the thermal flux distribution in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

    1986-01-01

    The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

  5. Computer-aided testing and operational aids for PARR-1 nuclear reactor

    International Nuclear Information System (INIS)

    Ansari, S.A.

    1990-01-01

    The utilization of the plant computer of Pakistan Research Reactor (PARR-1) for automatic periodic testing of nuclear instrumentation in the reactor is described. Computer algorithms have been developed for on-line acquisition and real-time processing of nuclear channel signals. The mean value, standard deviation, and probability distributions of nuclear channel signals are obtained in real time, and the computer generates a warning message if the signal error exceeds the maximum permissible error. In this way a faulty channel is automatically identified. Other real-time algorithms are also described that assist the operator in safe reactor operation by automatically computing approach-to-criticality during reactor start-up and the control rod worth determination

  6. Properties of autoregressive model in reactor noise analysis, 1

    International Nuclear Information System (INIS)

    Yamada, Sumasu; Kishida, Kuniharu; Bekki, Keisuke.

    1987-01-01

    Under appropriate conditions, stochastic processes are described by the ARMA model, however, the AR model is popularly used in reactor noise analysis. Hence, the properties of AR model as an approximate representation of the ARMA model should be made clear. Here, convergence of AR-parameters and PSD of AR model were studied through numerical analysis on specific examples such as the neutron noise in subcritical reactors, and it was found that : (1) The convergence of AR-parameters and AR model PSD is governed by the ''zero nearest to the unit circle in the complex plane'' (μ -1 ,|μ| M . (3) The AR model of the neutron noise of subcritical reactors needs a large model order because of an ARMA-zero very close to unity corresponding to the decay constant of the 6-th group of delayed neutron precursors. (4) In applying AR model for system identification, much attention has to be paid to a priori unknown error as an approximate representation of the ARMA model in addition to the statistical errors. (author)

  7. SCALE-4 analysis of pressurized water reactor critical configurations. Volume 1: Summary

    International Nuclear Information System (INIS)

    DeHart, M.D.

    1995-03-01

    The requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. If credit is to be taken for the reduced reactivity of burned or spent fuel relative to its original fresh composition, it is necessary to benchmark computational methods used in determining such reactivity worth against spent fuel reactivity measurements. This report summarizes a portion of the ongoing effort to benchmark away-from-reactor criticality analysis methods using critical configurations from commercial pressurized water reactors (PWR). The analysis methodology utilized for all calculations in this report is based on the modules and data associated with the SCALE-4 code system. Each of the five volumes comprising this report provides an overview of the methodology applied. Subsequent volumes also describe in detail the approach taken in performing criticality calculations for these PWR configurations: Volume 2 describes criticality calculations for the Tennessee Valley Authority's Sequoyah Unit 2 reactor for Cycle 3; Volume 3 documents the analysis of Virginia Power's Surry Unit 1 reactor for the Cycle 2 core; Volume 4 documents the calculations performed based on GPU Nuclear Corporation's Three Mile Island Unit 1 Cycle 5 core; and, lastly, Volume 5 describes the analysis of Virginia Power's North Anna Unit 1 Cycle 5 core. Each of the reactor-specific volumes provides the details of calculations performed to determine the effective multiplication factor for each reactor core for one or more critical configurations using the SCALE-4 system; these results are summarized in this volume. Differences between the core designs and their possible impact on the criticality calculations are also discussed. Finally, results are presented for additional analyses performed to verify that solutions were sufficiently converged

  8. Spent fuel management plans for the FiR 1 Reactor

    International Nuclear Information System (INIS)

    Salmenhaara, S. E. J.

    2002-01-01

    The FiR 1-reactor, a 250 kW TRIGA reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. The final disposal site is situated in Olkiluoto, on the western coast of Finland. Olkiluoto is also one of the two nuclear power plant sites in Finland. In the new operating license of our reactor there is a special condition. We have to achieve a binding agreement between our Research Centre and either the domestic Nuclear Power Companies about the possibility to use the Olkiluoto final disposal facility for our spent fuel or US DOE about the return of our spent fuel back to USA. If we want to continue the reactor operation beyond the year 2006. the domestic final disposal is the only possibility. At the moment it seems to be reasonable to prepare to both possibilities: the domestic final disposal and the return to the USA offered by US DOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will decide, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNCT treatments will cover the costs. If the BNCT and other irradiations develop satisfactorily, the reactor can be kept in operation beyond the year 2006 and the domestic final disposal will be implemented. If, however, there is still lack of money, there is no reason to continue the operation of the reactor and the choice of US DOE alternative is natural. (author)

  9. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    International Nuclear Information System (INIS)

    Mesquita, Henrique F.A.; Ferreira, Andrea V.

    2015-01-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  10. Irradiation experience of IPEN fuel at IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Perrotta, Jose A.; Neto, Adolfo; Durazzo, Michelangelo; Souza, Jose A.B. de; Frajndlich, Roberto

    1998-01-01

    IPEN/CNEN-SP produces, for its IEA-R1 Research Reactor, MTR fuel assemblies based on U 3 O 8 -Al dispersion fuel type. Since 1985 a qualification program on these fuel assemblies has been performed. Average 235 U burnup of 30% and peak burnup of 50% was already achieved by these fuel assemblies. This paper presents some results acquire, by these fuel assemblies, under irradiation at IEA-R1 Research Reactor. (author)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. Possible future roles for fast breeder reactors Part 1 and 2

    International Nuclear Information System (INIS)

    1978-06-01

    Part 1. The Fast Breeder Reactor (in particular in its sodium cooled version) has been steadily developed in the Community. This report attempts to quantify the advantages of this system in terms of fossil energy and uranium savings in the medium/long term as well as to examine some long term economic implications. The methodology of comparing scenarios, not individual reactor systems is followed. These scenarios have been chosen taking into account a range of assumptions concerning Community energy demand growth, fossil energy and uranium availability and technological capabilities. Part 2. The fast breeder reactor (FBR), particularly its sodium-cooled form (LMFBR) has been under development in the Community for many years. Industrial enterprises dedicated to its commercialisation have been formed and long range plans for its industrial utilisation are being formulated. The value of breeder reactors from the point of view of minimising Community fuel requirements has been discussed in Part I of this report (1). In Part II the consequences of delaying their introduction, and the demands placed upon the recycle industry by the introduction of fast reactors of different characteristics, using the Community electricity demand scenarios developed for Part I, are discussed. In addition comments are provided upon the effect of FBR introduction on the size of plutonium stocks

  13. The AMPS 1.5 MW low-pressure compact reactor

    International Nuclear Information System (INIS)

    Hewitt, J.S.

    1987-01-01

    The 1.5-MWt reactor of the Autonomous Marine Power Source (AMPS) is designed to meet the unusual requirements of its first application. To provide for 100 kWe (net) on board self-sustaining manned submersible vehicles, the AMPS reactor must deliver safely, reliably and without direct operator surveillance, its thermal output to freon Rankine-cycle engines at thermodynamically useful temperatures. It must also conform to space and weight limits on the order of less than 50 cubic metres and 70 tonnes. The safety requirements are met by (i) limiting lifetime excess reactivity requirements by incorporation of burnable poison in the U-Zr-H fuel, (ii) maintaining nominal pressures in the light-water primary system at about 1 atmosphere, and (iii) maintaining a large volume of primary reserve coolant at temperature depressed relative to that of the circulating coolant. The latter averages 90 degrees celsius as it is pumped around loops that include the reactor core and the freon evaporators during normal operation. In the event of loss of pumped flow, the system defaults by intrinsic means to core cooling through natural convective exchange with the reserve coolant. In the post-shutdown situation, this passive cooling mode continues to operate regardless of vessel orientation and decay heat is safely dissipated to the sea. The design of the AMPS system, including the reactor, the freon engines, the control and monitoring system, the safety shut-down system and the power source container, are in advanced stages of design. (author)

  14. Plasma Physics Research Institute, Lawrence Livermore National Laboratory, University of California, Davis annual report for fiscal year 1989

    International Nuclear Information System (INIS)

    Killeen, J.; Drake, R.P.

    1991-01-01

    This report discusses: The Davis Diverted Tokamak; Particle Simulation of Transport in Fusion Devices; Astrophysical Plasmas; Statistical Dynamics of Multi-Field Models for Plasma; Large Scale Density Modifications Induced in the Ionosphere; Studies of the Ion Acoustic Decay Instability; and Computer Simulation of Ionospheric Radio Frequency Heating

  15. Flux distribution measurements in the Bruce A unit 1 reactor

    International Nuclear Information System (INIS)

    Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

    1977-07-01

    Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

  16. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2002-07-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  17. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    Science.gov (United States)

    El-Genk, Mohamed S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept.

  18. Pellet bed reactor for nuclear propelled vehicles: Part 1: Reactor technology

    International Nuclear Information System (INIS)

    El-genk, M.S.

    1991-01-01

    The pellet bed reactor (PBR) for nuclear propelled vehicles is briefly discussed. Much of the information is given in viewgraph form. Viewgraphs include information on the layout for a Mars mission using a PBR nuclear thermal rocket, the rocket reactor layout, the fuel pellet design, materials compatibility, fuel microspheres, microsphere coating, melting points in quasibinary systems, stress analysis of microspheres, safety features, and advantages of the PBR concept

  19. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lightwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. The research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology, are presented. (Author) [pt

  20. Experience and research with the IEA-R1 Brazilian reactor

    International Nuclear Information System (INIS)

    Fulfaro, R.; Sousa, J.A. de; Nastasi, M.J.C.; Vinhas, L.A.; Lima, F.W. de.

    1982-06-01

    The IEA-R1 reactor of the Instituto de Pesquisas Energeticas e Nucleares, IPEN, of Sao Paulo, Brazil, a lighwater moderated swimming-pool research reactor of MTR type, went critical for the first time on September 16, 1957. In a general way, in these twenty four years the reactor was utilized without interruption by users of IPEN and other institutions, for the accomplishment of work in the field of applied and basic research, for master and doctoral thesis and for technical development. Some works performed and the renewal programme established for the IEA-R1 research reactor in which several improvements and changes were made. Recent activities in terms of production of radioisotopes and some current research programm in the field of Radiochemistry are described, mainly studies and research on chemical reactions and processes using radioactive tracers and development of radioanalytical methods, such as neutron activation and isotopic dilution. It is also presented the research programmes of the Nuclear Physics Division of IPEN, which includes: nuclear spectroscopy studies and electromagnetic hyperfine interactions; neutron diffraction; neutron inelastic scattering studies in condensed matter; development and application of the technique of fission track register in solid state detectors; neutron radioactive capture with prompt gamma detection and, finally, research in the field of nuclear metrology. (Author) [pt

  1. Diversity of Rhynchosporium secalis (Oud. J. J. Davis strains in morphological and cultural peculiarities

    Directory of Open Access Journals (Sweden)

    L. Lebedeva

    2012-12-01

    Full Text Available Biological peculiarities of the rye scald fungus Rhynchosporium secalis (Oud. J. J. Davis, in one population of North-West region were examined. Seventy-eight isolates, the causal agent of scald, were taken from infected rye plants. This isolates were analalysed on rate of growth on artificial test medium, structure and color and temperature dependence. Single-spore strains were obtained from each natural isolate. Color and structure of some single-spore isolates remained stable through repeated transfers to fresh PDA medium.

  2. CAC-RA1 1958-1998. The first years of the Constituyentes Atomic Center (CAC). History of the first Argentine nuclear reactor (RA-1); CAC-RA-1 1958-1998. Los primeros anios del CAC. Historia del primer reactor nuclear argentino (RA-1)

    Energy Technology Data Exchange (ETDEWEB)

    Forlerer, Elena; Palacios, Tulio A [comps.

    1998-07-01

    After giving the milestones of the development of the Constituyentes Atomic Center since 1957, the history of the construction of the first nuclear reactor (RA-1) in Argentina, including the local fabrication of its fuel elements, is surveyed. The RA-1 reached criticality on January 17, 1958. The booklet commemorates the 40th year of the reactor operation.

  3. Benchmark testing of Canadol-2.1 for heavy water reactor

    International Nuclear Information System (INIS)

    Liu Ping

    1999-01-01

    The new version evaluated nuclear data library of ENDF-B 6.5 has been released recently. In order to compare the quality of evaluated nuclear data CENDL-2.1 with ENDF-B 6.5, it is necessary to do benchmarks testing for them. In this work, CENDL-2.1 and ENDF-B 6.5 were used to generated the WIMS-69 group library respectively, and benchmarks testing was done for the heavy water reactor, using WIMS5A code. It is obvious that data files of CENDL-2.1 is better than that of old WIMS library for the heavy water reactors calculations, and is in good agreement with those of ENDF-B 6.5

  4. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  5. Use of self powered neutron detectors in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Galo Rocha, F. del.

    1989-01-01

    A survey of self-powered neutron detectors, SPND, which are used as part of the in-core instrumentation of nuclear reactors is presented. Measurements with Co and Er SPND's were made in the IEA-R1 reactor for determining the neutron flux distribution and the integral reactor power. Due to the size of the available detectors, the neutron flux distribution could not be obtained with accuracy. The results obtained in the reactor power measurements demonstrate that the SPND have the linearity and the quick response necessary for a reactor power channel. This work also presents a proposed design of a SPND using Pt as wire emissor. This proposed design is based in the experience gained in building two prototypes. The greatest difficulties encountered include materials and technology to perform the delicate weldings. (author)

  6. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    Energy Technology Data Exchange (ETDEWEB)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn [Physics Division, Office of Atomic Energy for Peace, Vibhavadi Rangsit Road, Chatuchak, Bangkok (Thailand)

    1999-08-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10{sup 7} n.cm{sup 2}.sec{sup -1} at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  7. Status of prompt gamma neutron activation analysis (PGAA) at TRR-1/M1 (Thai Research Reactor-1/Modified 1)

    International Nuclear Information System (INIS)

    Asvavijnijkulchai, Chanchai; Dharmavanij, Wanchai; Siangsanan, Pariwat; Ratanathongchai, Wichian; Chongkum, Somporn

    1999-01-01

    The first prompt gamma activation analysis (PGAA) was designed, constructed and installed at a 6 inch diameter neutron beam port of the Thai Research Reactor-1/Modified 1 (TRR-1/M1) since 1989. Beam characteristic were made by Gd foil irradiation, X-ray film exposing and densitometry scanning consequently. The thermal neutron flux at sample position was measured by Au foil activation, and was about 1 x 10 7 n.cm 2 .sec -1 at 700 kW operating power. The experiments have been conducted successfully. In 1998, the PGAA facility has been developed for the reactor operating power at 1.2 MW. The new PGAA system, e.g., beam shutter, gamma collimator and biological shields have been designed to reduce the leakage of neutrons and gamma radiation to the acceptance levels in accordance with the International Commission on Radiation Protection Publication 60 (ICRP 60). The construction and installation will be completed in April 1999. (author)

  8. Reactor inventory monitoring system for Angra-1 reactor

    International Nuclear Information System (INIS)

    S Neto, Joaquim A.; Silva, Marcos C.; Pinheiro, Ronaldo F.M.; Soares, Milton; Martinez, Aquilino; Comerlato, Cesar A.; Oliveira, Eugenio A.

    1996-01-01

    This work describes the project of Reactor Inventory Monitoring System, which will be installed in Angra I Nuclear Power Plant. The inventory information is important to the operators take corrective actions in case of an incident that may cause a failure in the core cooling. (author)

  9. Modifications done in the IPR-R1 reactor and their auxiliary systems

    International Nuclear Information System (INIS)

    Maretti Junior, F.; Amorim, V.A. de; Coura, J.G.

    1986-01-01

    The improvements done in the IPR-R1 reactor for adequateness of operation conditions and increase of irradiation sample capability. The cooling systems, reactor pool, system of control rods were substituted. The optimization of transfer pneumatic system was done. (M.C.K.) [pt

  10. John Davies of Hereford, the King of Denmark & Shakespeare's Meeting of Kings: Praise Beyond Praise

    DEFF Research Database (Denmark)

    Sterrett, Joseph William

    2016-01-01

    This article traces the response and style of John Davies of Hereford, 'an ordinary man' as he celebrated an extraordinary event, the state visit of the King of Denmark to the court of James I in 1606. It then draws comparisons to Shakespeare's meeting of kings some seven or eight years later...... at the beginning of the late history play, All is True, suggesting that the earlier poet's experience influenced the latter....

  11. Proceedings of 2. Yugoslav symposium on reactor physics, Part 1, Herceg Novi (Yugoslavia), 27-29 Sep 1966

    International Nuclear Information System (INIS)

    1966-01-01

    This Volume 1 of the Proceedings of 2. Yugoslav symposium on reactor physics includes nine papers dealing with the following topics: reactor kinetics, reactor noise, neutron detection, methods for calculating neutron flux spatial and time dependence in the reactor cores of both heavy and light water moderated experimental reactors, calculation of reactor lattice parameters, reactor instrumentation, reactor monitoring systems; measuring methods of reactor parameters; reactor experimental facilities

  12. William Morris Davis e a Teoria Geográfica

    Directory of Open Access Journals (Sweden)

    Carlos Augusto de Figueiredo Monteiro

    2001-12-01

    Full Text Available O autor põe em confronto a proposta teórica do Ciclo Geográfico de W. M. Davis, da virada dos séculos passados (1899 e a famosa crítica sobre a carência de fundamentação científica na Geografia, feita por Fred Schaefer (1953. Este é o ponto de partida para traçar um panorama da evolução da Geografia Física no Brasil, notadamente da Geomorfologia e da Climatologia, ao longo do século XX. Sincronizando a evolução do pensamento geográfico com os grandes acontecimentos mundiais do passado século, o autor destaca o segmento 1968-1973 como o possível ponto de mutação a partir do qual penetramos na Grande Crise Histórica que atravessamos nesta virada dos séculos XX e XXI e o caráter desagregativo da atual Geografia.

  13. Thai research reactor

    International Nuclear Information System (INIS)

    Aramrattana, M.

    1987-01-01

    The Office of Atomic Energy for Peace (OAEP) was established in 1962, as a reactor center, by the virtue of the Atomic Energy for Peace Act, under operational policy and authority of the Thai Atomic Energy for Peace Commission (TAEPC); and under administration of Ministry of Science, Technology and Energy. It owns and operates the only Thai Research Reactor (TRR-1/M1). The TRR-1/M1 is a mixed reactor system constituting of the old MTR type swimming pool, irradiation facilities and cooling system; and TRIGA Mark III core and control instrumentation. The general performance of TRR-1/M1 is summarized in Table I. The safe operation of TRR-1/M1 is regulated by Reactor Safety Committee (RSC), established under TAEPC, and Health Physics Group of OAEP. The RCS has responsibility and duty to review of and make recommendations on Reactor Standing Orders, Reactor Operation Procedures, Reactor Core Loading and Requests for Reactor Experiments. In addition,there also exist of Emergency Procedures which is administered by OAEP. The Reactor Operation Procedures constitute of reactor operating procedures, system operating procedures and reactor maintenance procedures. At the level of reactor routine operating procedures, there is a set of Specifications on Safety and Operation Limits and Code of Practice from which reactor shift supervisor and operators must follow in order to assure the safe operation of TRR-1/M1. Table II is the summary of such specifications. The OAEP is now upgrading certain major components of the TRR-1/M1 such as the cooling system, the ventilation system and monitoring equipment to ensure their adequately safe and reliable performance under normal and emergency conditions. Furthermore, the International Atomic Energy Agency has been providing assistance in areas of operation and maintenance and safety analysis. (author)

  14. Continuous backfitting measures for the FRG-1 and FRG-2 research reactors

    International Nuclear Information System (INIS)

    Blom, K.H.; Falck, K.; Krull, W.

    1990-01-01

    The GKSS-Research Centre Geesthacht GmbH has been operating the research reactors FRG-1 and FRG-2 with power levels of 5 MW and 15 MW for 31 and 26 years respectively. Safe operation at full power levels over so many years with an average utilization between 180 d to 250 d per year is possible only with great efforts in modernization and upgrading of the research reactors. Emphasis has been placed on backfitting since around 1975. At that time within the Federal Republic of Germany many new guidelines, rules, ordinances, and standards in the field of (power) reactor safety were published. Much work has been done on the modernization of the FRG-1 and FRG-2 research reactors therefore within the last ten years. Work done within the last two years and presently underway includes: measures against water leakage through the concrete and along the beam tubes; repair of both cooling towers; modernization of the ventilation system; measures for fire protection; activities in water chemistry and water quality; installation of a double tubing for parts of the primary piping of the FRG-1; replacement of instrumentation, process control systems (operation and monitoring system) and alarm system; renewal of the emergency power supply; installation of internal lightning protection; installation of a cold neutron source; enrichment reduction for FRG-1. These efforts will continue to allow safe operation of our research reactors over their whole operational life

  15. Instrumentation renewal at the FIR 1 research reactor in Finland

    International Nuclear Information System (INIS)

    Bars, Bruno; Kall, Leif

    1982-01-01

    The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

  16. Dose measurements in controlled area of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, F.L.; Junior, F.M.

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers

  17. Demolition of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklearservice, Essen (Germany); Cremer, J. [SNT Siempelkamp Nukleartechnik, Heidelberg (Germany)

    2003-06-01

    FRJ-2 (MERLIN), the swimming pool reactor cooled and moderated by light water, was built at the then Juelich Nuclear Research Establishment (KFA) between 1958 and 1962. In the period between 1964 and 1985, it was used for. The reactor was decommissioned in 1985. Since 1996, most of the demolition work has been carried out under the leadership of a project team. The complete secondary cooling system was removed by late 1998. After the cooling loops and experimental installations had been taken out, the reactor vessel internals were removed in 2000 after the water had been drained from the reactor vessel. After the competent authority had granted a license, demolition of the reactor block, the central part of the research reactor, was begun in October 2001. In a first step, the reactor operating floor and the reactor attachment structures were removed by the GNS/SNT consortium charged with overall planning and execution of the job. This phase gave rise to approx. The reactor block proper is dismantled in a number of steps. A variety of proven cutting techniques are used for this purpose. Demolition of the reactor block is to be completed in the first half of 2003. (orig.) [German] Der mit Leichtwasser gekuehlte und moderierte Schwimmbad-Forschungsreaktor FRJ-2 (MERLIN) wurde von 1958 bis 1962 fuer die damalige Kernforschungsanlage Juelich (KFA) errichtet. Von 1964 bis 1985 wurde er fuer Experimente mit zunaechst 5 MW und spaeter 10 MW thermischer Leistung bei einem maximalen thermischen Neutronenfluss von 1,1.10{sup 14} n/cm{sup 2}s genutzt. Im Jahr 1985 stellte der Reaktor seinen Betrieb ein. Die Brennelemente wurden aus der Anlage entfernt und in die USA und nach Grossbritannien verbracht. Seit 1996 erfolgen die wesentlichen Abbautaetigkeiten unter Leitung eines verantwortlichen Projektteams. Bis Ende 1998 wurde das komplette Sekundaerkuehlsystem entfernt. Dem Abbau der Kuehlkreislaeufe und Experimentiereinrichtungen folgte im Jahr 2000 der Ausbau der

  18. Equipment for neutron measurements at VR-1 Sparrow training reactor

    International Nuclear Information System (INIS)

    Kolros, Antonin; Huml, Ondrej; Kos, Josef

    2008-01-01

    Full text: The VR-1 Sparrow training reactor is the experimental nuclear facility especially employed for education and teaching of students from different technical universities in the Czech Republic and other countries. Since 2005 the uniform all-purpose devices EMK310 have been used for measurement at reactor laboratory with different type of gas filled neutron detectors. The neutron detection system are employed for reactivity measurement, control rod calibration, critical experiment, study of delayed neutrons, study of nuclear reactor dynamics and study of detection systems dead time. The small dimension isotropic detectors are especially used for measurement of thermal neutron flux distribution inside the reactor core. The EMK-310 is a high performance, portable, three-channel fast amplitude analyzer designed for counting applications. It was developed for nuclear applications and made in close co-operation with firm TEMA Ltd. The precise rack eliminates electromagnetic disturbance and contains the control unit and four modules. The modules of high voltage supply and amplifier for gas filled detectors or scintillation probes are used in basic configuration. Software is tailored specifically to the reactor measurement and allows full online control. For applications involving the study of signals that may vary with the time, example study of delayed neutrons or nuclear reactor dynamics, the EMK-310 provides a Multichannel Scaling (MCS) acquisition mode. MCS dwell time can be set from 2 ms. Now, the new generation of digital multichannel analyzers DA310 is introduced. They have similarly attributes as EMK310 but the output information of unipolar signals from detector is more complete. The pipeline A/D converter with field programmable gate array (FPGA) is the hearth of the DA310 device. The resolution is 12 bits (4096 channels); the sample frequency is 80 MHz. The application for the neutron noise analysis is supposed. The correction method for non linearity

  19. Report on safety related occurrences and reactor trips July 1, 1979 - December 31, 1979

    International Nuclear Information System (INIS)

    Olsson, S.; Andermo, L.

    1980-01-01

    This is a report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1979 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 76 safety related occurrences and 27 reactor trips have been reported to the Nuclear Power Inspectorate. It is to the greatest extent conventional components such as valves and pumps which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant system and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips are normal. The average value for these 6 months is 4.5 trips/unit. Approximetely one half of the reactor trips happened at zero or very low power operation. The fact that even small deviations from prescribed operation result in an automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences without safety significance. (author)

  20. Normative Data for the Balance Error Scoring System in Adults

    Directory of Open Access Journals (Sweden)

    Grant L. Iverson

    2013-01-01

    Full Text Available Background. The balance error scoring system (BESS is a brief, easily administered test of static balance. The purpose of this study is to develop normative data for this test. Study Design. Cross-sectional, descriptive, and cohort design. Methods. The sample was drawn from a population of clients taking part in a comprehensive preventive health screen at a multidisciplinary healthcare center. Community-dwelling adults aged 20–69 (N=1,236 were administered the BESS within the context of a fitness evaluation. They did not have significant medical, neurological, or lower extremity problems that might have an adverse effect on balance. Results. There was a significant positive correlation between BESS scores and age (r=.34. BESS performance was similar for participants between the ages of 20 and 49 and significantly declined between ages 50 and 69. Men performed slightly better than women on the BESS. Women who were overweight performed significantly more poorly on the test compared to women who were not overweight (P<.0001; Cohen's d=.62. The BESS normative data are stratified by age and sex. Conclusions. These normative data provide a frame of reference for interpreting BESS performance in adults who sustain traumatic brain injuries and adults with diverse neurological or vestibular problems.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  2. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  4. IPR-R1 TRIGA research reactor decommissioning plan

    International Nuclear Information System (INIS)

    Andrade Grossi, Pablo; Oliveira de Tello, Cledola Cassia; Mesquita, Amir Zacarias

    2008-01-01

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  5. The Chernobyl reactor accident. Pt. 1 and 2

    International Nuclear Information System (INIS)

    1986-06-01

    The report first summarizes the available information on the various incidents of the whole accident scenario, and combines the information to present a first general outline and a basis for appraisal. The most significant incidents reported, namely power excursion, core meltdown, and fire, are discussed with a view to the reactor design and safety of reactors installed in the FRG. The main differences and advantages of German reactor designs are shown, as e.g.: Power excursions are mastered by inherent physical conditions; far better redundancy of engineered safety systems; enclosure of the complete reactor cooling system in a pressure-retaining steel containment; reactor buildings being made of reinforced concrete. The second part of the report deals with the radiological effects to be expected for our country. Data are given on the varying radiological exposure of the different regions. The fate and uptake of radioactivity in the human body are discussed. The conclusion drawn from the data presented is that the individual exposure due to the reactor accident will remain within the variations and limits of natural radioactivity and effects. (orig./HP) [de

  6. Final Environmental Assessment for a Solar Power System at Davis-Monthan Air Force Tucson, Arizona

    Science.gov (United States)

    2009-09-01

    to the following factors depending on the corresponding years. Year 2005 through 2009: VOCE = .016 * Trips NOxE = .015 * Trips PM10E = .0022...Trips COE = .262 * Trips Year 2010 and beyond: VOCE = .012 * Trips NOxE = .013 * Trips PM10E = .0022 * Trips COE = .262 * Trips FINAL...ENVIRONMENTAL ASSESSMENT B-8 Solar Power System (SPS) at Davis-Monthan AFB To convert from pounds per day to tons per year: VOC (tons/yr) = VOCE * DPYII/2000

  7. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    2002-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  8. Regulatory approach to enhanced human performance during accidents

    International Nuclear Information System (INIS)

    Palla, R.L. Jr.

    1990-01-01

    It has become increasingly clear in recent years that the risk associated with nuclear power is driven by human performance. Although human errors have contributed heavily to the two core-melt events that have occurred at power reactors, effective performance during an event can also prevent a degraded situation from progressing to a more serious accident, as in the loss-of-feedwater event at Davis-Besse. Sensitivity studies in which human error rates for various categories of errors in a probabilistic risk assessment (PRA) were varied confirm the importance of human performance. Moreover, these studies suggest that actions taken during an accident are at least as important as errors that occur prior to an initiating event. A program that will lead to enhanced accident management capabilities in the nuclear industry is being developed by the US Nuclear Regulatory Commission (NRC) and industry and is a key element in NRC's integration plan for closure of severe-accident issues. The focus of the accident management (AM) program is on human performance during accidents, with emphasis on in-plant response. The AM program extends the defense-in-depth principle to plant operating staff. The goal is to take advantage of existing plant equipment and operator skills and creativity to find ways to terminate accidents that are beyond the design basis. The purpose of this paper is to describe the NRC's objectives and approach in AM as well as to discuss several human performance issues that are central to AM

  9. Insertion of reactivity (RIA) without scram in the reactor core IEA-R1 using code PARET

    International Nuclear Information System (INIS)

    Alves, Urias F.; Castrillo, Lazara S.; Lima, Fernando A.

    2013-01-01

    The modeling and analysis thermo hydraulics of a research reactor with MTR type fuel elements - Material Testing Reactor - was performed using the code PARET (Program for the Analysis of Reactor Transients) when in the system some external event is introduced that changed the reactivity in the reactor core. Transients of Reactivity Insertion of 0.5 , 1.5 and 2.0$/ 0.7s in the brazilian reactor IEA-R1 will be presented, and will be shown under what conditions it is possible to ensure the safe operation of its nucleus. (author)

  10. Expanding the storage capability at ET-RR-1 research reactor at Inshass

    International Nuclear Information System (INIS)

    Sultan, Mariy M.; Khattab, M.

    1999-01-01

    Storing of spent fuel from Test Reactor in developing countries has become a big dilemma for the following reasons: The transportation of spent fuel is very expensive; There are no reprocessing plants in most developing countries; The expanding of existing storage facilities in reactor building require experience that most of developing countries lack; Some political motivations from Nuclear Developed countries intervene which makes the transportation procedures and logistics to those countries difficult. This paper gives the conceptual design of a new spent fuel storage now under construction at Inshass research reactor (ET-RR-1). The location of the new storage facility is chosen to be within the premises of the reactor facility so that both reactor and the new storage are one Material Balance Area. The paper also proposes some ideas that can enhance the transportation and storage of spent fuel of test reactors, such as: Intensifying the role of IAEA in helping countries to get rid of the spent fuel; The initiation of regional spent fuel storage facilities in some developing countries. (author)

  11. Licensing of the first reload of Angra-1 reactor

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.

    1985-01-01

    The historical aspects related to the licensing of the first reload of Angra-1 reactor are presented. The dates, the institutions, the experts, as well as the documents generated during that process are presented. (M.I.)

  12. Evaluation of the trial design studies for an advanced marine reactor, (1)

    International Nuclear Information System (INIS)

    1988-03-01

    The trial design of three type reactors, semi-integrated, integrated and integrated (self-pressurized) type, was carried out in order to clarify the reactor type for the advanced marine reactor that would be developed for its realization in future and in order to extract its research and development theme. The trial design was carried and finished as for the three type reactors in same specifications in order to improve the following characteristics, small in size, light in weight, high in safety and reliability, and economic. In this report, a comparison and review of the following items are described as for the above three type reactors, (1) specifications, (2) shielding, (3) refueling, (4) in-service inspection, (5) analysis of the transients and accidents, (6) piping systems, (7) control systems, (8) dynamic analysis, (9) overall comparison, (10) research and development theme and theme for study in future. (author)

  13. Transformation of 1,1,1-trichloroethane in an anaerobic packed-bed reactor at various concentrations of 1,1,1-trichloroethane, acetate and sulfate

    NARCIS (Netherlands)

    deBest, JH; Jongema, H; Weijling, A; Doddema, HJ; Janssen, DB; Harder, W

    Biotransformation of 1,1,1-trichloroethane (CH3CCl3) was observed in an anaerobic packed-bed reactor under conditions of both sulfate reduction and methanogenesis. Acetate (1 mM) served as an electron donor. CH3CCl3 was completely converted up to the highest investigated concentration of 10 mu M.

  14. Evaluation of power behavior during startup and shutdown procedures of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante M.; Mesquita, Amir Z.; Ferreira, Andrea V.

    2009-01-01

    The IPR-R1 nuclear reactor of Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN is a TRIGA Mark I pool type reactor cooled by natural circulation of light water. In the IPR-R1, the power is measured by four nuclear channels, neutron-sensitive chambers, which are mounted around the reactor core: the Startup Channel for power indication during reactor startup; the Logarithmic Wide Range Power Monitoring Channel; the Linear Multi-Range Power Monitoring Channel and the Percent Power Safety Channel. A data acquisition system automatically does the monitoring and storage of all the reactor operational parameters including the reactor power. The startup procedure is manual and the time to reach the desired reactor power level is different on each irradiation which may introduces differences in induced activity of samples irradiated in different irradiations. In this work, the power evolution during startup and shutdown periods of IPR-R1 operation was evaluated and the mean values of reactor energy production in these operational phases were obtained. The analyses were performed on basis of the Linear Multi-Range Channel data. The results show that the sum of startup and shutdown periods corresponds to 1% of released energy for irradiations during 1h at 100kW. This value may be useful to correct experimental data in neutron activation experiments. (author)

  15. Modernization of Safety and Control Instrumentation of the IEA-R1 Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    De Carvalho, P.V., E-mail: paulov@ien.gov.br [Institute of Nuclear Engineering (IEN), National Nuclear Energy Commission (CNEN), Rio de Janeiro (Brazil)

    2014-08-15

    The research reactor IEA-R1 located in the Institute of Energy and Nuclear Research (IPEN), São Paulo, Brazil, obtained its first criticality on 16 September 1957 and since then has served the scientific and medical community in the performance of experiments in applied nuclear physics, as well as the provision of radioisotopes for production of radiopharmaceuticals. The reactor produces radioisotopes {sup 82}Br and {sup 41}Ar for special processes in industrial inspection and {sup 192}Ir and {sup 198}Au as sources of radiation used in brachytherapy, {sup 153}Sm for pain relief in patients with bone metastasis, and calibrated sources of {sup 133}Ba, {sup 137}Cs, {sup 57}Co, {sup 60}Co, {sup 241}Am and {sup 152}Eu used in medical clinics and hospitals practicing nuclear medicine and research laboratories. Services are offered in regular non-destructive testing by neutron radiography, neutron irradiation of silicon for phosphorous doping and other various irradiations with neutrons. The reactor is responsible for producing approximately 70% of radiopharmaceutical {sup 131}I used in Brazil, which saves about US$ 800 000 annually for the country. After more than 50 years of use, most of its equipment and systems have been modernized, and recently the reactor power was increased to 5 MW in order to enhance radioisotope production capability. However, the control room and nuclear instrumentation system used for reactor safety have operated more than 30 years and require constant maintenance. Many equipment and electronic components are obsolete, and replacements are not available in the market. The modernization of the nuclear safety and control instrumentation systems of IEA-R1 is being carried out with consideration for the internationally recognized criteria for safety and reliable reactor operations and the latest developments in nuclear electronic technology. The project for the new reactor instrumentation system specifies three wide range neutron monitoring

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

    International Nuclear Information System (INIS)

    Martinc, R.; Stanic, A.

    1985-01-01

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

  17. Report on safety related occurrences and reactor trips July 1, 1977 - December 31, 1977

    International Nuclear Information System (INIS)

    Andermo, L.; Sundman, B.

    1974-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1 to December 31, 1977 inclusive. The facilities involved are Barsebaeck 1 and 2, Oskarshamn 1 and 2 and Ringhals 1 and 2. During this period of 6 months 48 safety related occurrences and 49 reactor trips have been reported to the Nuclear Power Inspectorate. Included is also one incident June 21 in Barsebaeck 2 which was not included in the last compilation of occurrences. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The total number of reactor trips have increased nearly 30% since the last period. Those occurred during power operation however, were less. More than 50% of the reactor trips happened in the shutdown condition. The fact that even small deviations from prescribed operation result in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The greatest outages are caused by occurrences withou02068NRM 0000169 450

  18. NRDA-processed CTD data from the HOS Davis in the Gulf of Mexico, Cruise 3 Leg 1, collected from 2010-09-09 to 2010-09-27, associated with the Deepwater Horizon Oil Spill event (NCEI Accession 0130017)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Conductivity Temperature and Depth (CTD) measurements were collected aboard the R/V HOS Davis, Cruise 03, to determine physical oceanographic parameters of the water...

  19. NRDA-processed CTD data from the HOS Davis in the Gulf of Mexico, Cruise 4 Leg 1, collected from 2010-11-07 to 2010-11-14, associated with the Deepwater Horizon Oil Spill event (NCEI Accession 0130023)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Conductivity Temperature and Depth (CTD) measurements were collected aboard the R/V HOS Davis, Cruise 04, to determine physical oceanographic parameters of the water...

  20. NRDA-processed CTD data from the HOS Davis in the Gulf of Mexico, Cruise 2 Leg 1, collected from 2010-08-26 to 2010-09-02, associated with the Deepwater Horizon Oil Spill event (NCEI Accession 0128093)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Conductivity Temperature and Depth (CTD) measurements were collected aboard the R/V HOS Davis, Cruise 02, to determine physical oceanographic parameters of the water...

  1. Application of a quantitative histological health index for Antarctic rock cod (Trematomus bernacchii) from Davis Station, East Antarctica.

    Science.gov (United States)

    Corbett, Patricia A; King, Catherine K; Mondon, Julie A

    2015-08-01

    A quantitative Histological Health Index (HHI) was applied to Antarctic rock cod (Trematomus bernacchii) using gill, liver, spleen, kidney and gonad to assess the impact of wastewater effluent from Davis Station, East Antarctica. A total of 120 fish were collected from 6 sites in the Prydz Bay region of East Antarctica at varying distances from the wastewater outfall. The HHI revealed a greater severity of alteration in fish at the wastewater outfall, which decreased stepwise with distance. Gill and liver displayed the greatest severity of alteration in fish occurring in close proximity to the wastewater outfall, showing severe and pronounced alteration respectively. Findings of the HHI add to a growing weight of evidence indicating that the current level of wastewater treatment at Davis Station is insufficient to prevent impact to the surrounding environment. The HHI for T. bernacchii developed in this study is recommended as a useful risk assessment tool for assessing in situ, sub-lethal impacts from station-derived contamination in coastal regions throughout Antarctica. Copyright © 2015 Elsevier Ltd. All rights reserved.

  2. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

    2015-09-01

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  3. Fumio Matsumura--accomplishments at the University of California, Davis, and in the Sierra Nevada Mountains.

    Science.gov (United States)

    Seiber, James N

    2015-05-01

    Fumio Matsumura joined the University of California, Davis, faculty in 1987 where he served as founding director of the Center for Environmental Health Sciences, associate director of the U.C. Toxic Substances Research and Teaching Program, and chair of the Department of Environmental Toxicology. He was an active affiliate with the NIEHS-funded Superfund Basic Research Program and the NIH Comprehensive Cancer Center. He was in many instances a primary driver or otherwise involved in most activities related to environmental toxicology at Davis, including the education of students in environmental biochemistry and ecotoxicology. A significant part of his broad research program was focused on the long range transport of chemicals such as toxaphene, PCBs and related contaminants used or released in California to the Sierra Nevada mountains, downwind of the urban and agricultural regions of the state. He hypothesized that these chemical residues adversely affected fish and wildlife, and particularly the declining populations of amphibians in Sierra Nevada streams and lakes. Fumio and his students and colleagues found residues of toxaphene and PCBs at higher elevations, an apparent result of atmospheric drift and deposition in the mountains. Fumio and his wife Teruko had personal interests in, and a love of the mountains, as avid skiers, hikers, and outdoor enthusiasts. Copyright © 2014 Elsevier Inc. All rights reserved.

  4. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  5. Report on safety related occurrences and reactor trips July 1, 1976-December 31, 1976

    International Nuclear Information System (INIS)

    Andermo, L.

    1977-04-01

    This is a systematically arranged report on all reported safety related occurrences and reactor trips in Swedish nuclear power plants in operation during July 1, 1976 to December 31, 1976 inclusive. The facilities involved are Oskarshamn 1 and 2, Ringhals 1 and 2 and Barsebaeck 1. During this period of the 6 months 37 safety related occurrences and 34 reactor trips have been reported to the Nuclear Power Inspectorate. As earlier experiences have shown it is to the greatest extent the conventional components which bring about the safety related occurrences or occurrences leading to outages or power reductions. However, the component errors discovered in the safety related systems have not affected the function of their redundant systems and other diverse systems have not been involved. Therefore the reactor safety has been satisfactory. The fact that even small deviations from prescribed operation results in automatic and safe shut down of the reactor, does not always imply a conflict with operational availability. The number of reactor trips are almost as low as during the last period, which is a drastic reduction compared to earlier time periods. The greatest outages are caused by occurrences without safety significance.(author)

  6. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    Energy Technology Data Exchange (ETDEWEB)

    El-Messeiry, A M [National Center for Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs.

  7. Probabilistic study of LOFA in ETRR-1 reactor. Vol. 4

    International Nuclear Information System (INIS)

    El-Messeiry, A.M.

    1996-01-01

    In evaluating the safety of a research reactor an analysis of reactor to a wide range of postulated initiating events must be carried out, that could lead to anticipated operational occurrences or accident conditions. These disturbances include decrease in heat removal by the reactor coolant system which may be due to loss of coolant flow (LOFA) or loss of coolant heat sink. LOFA is considered here for this study for the tank type research reactor with a probabilistic approach applied to (ET-RR-1). The reactor is provided with engineering safety system to respond to accidents and perform mitigating functions. The possible malfunctions, Failures, operator errors leading to LOFA initiating event are presented (pipe break; valve opening; pump failure ...etc.). The basic event frequency/probability is calculated using appropriate probability model. The logic event tree model is constructed to illustrate all possible accident scenarios. This scenario combines system success and failure probabilities with the probability of postulated initiating events occurring that result in an accident sequence probability associated with a certain plant state. Fault tree technique is adopted to determine engineering safety features probabilities. The results show the possible minimal cut sets of variable order of each system failure. Accident sequences leading to core damage state, effects of component failures, operator errors, and system failure on plant states. The possible weak points in the design are presented. 14 figs., 3 tabs

  8. Air pollutant emission rates for sources at the Davis Canyon Repository site

    International Nuclear Information System (INIS)

    1985-11-01

    This document summarizes the air-quality source terms used for the Davis Canyon, Utah environmental assessment report and explains their derivation. The engineering data supporting these source terms appear as appendixes to the report and include summary equipment lists for the repository (December, 1984) and detailed equipment lists for the exploratory shaft (June and July, 1985). Although substantial work has been performed in establishing the current repository design, a greater effort will be required for the final design. Consequently, the repository emission rates presented here should be considered as preliminary estimates. Another set of air pollutant emission rates will be calculated after design data are more firmly established. 19 refs., 18 tabs

  9. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  10. Paleocene Picrites of Davis Strait: Products of a Plume or Plates?

    Science.gov (United States)

    Beutel, E. K.; Clarke, D. B.

    2017-12-01

    Voluminous, subaerial, ultra-depleted, 62 Ma, primary picritic lavas occur on both sides of Davis Strait separating Baffin Island and West Greenland. Temporally, the picrites are coeval with the initiation of sea-floor spreading in Labrador Sea and Baffin Bay around 62 Ma. Petrogenetically, the chemical characteristics of these picrites (MgO = 18-21 wt. %; K2O = 0.01-0.20 wt. %; 87Sr/86Sri ≈ 0.7030; ɛNdi ≈ +5.2-8.6; 3He/4He ≤ 49.5RA) demand only derivation by partial melting of highly depleted subcontinental lithospheric mantle (SCLM) at a pressure of 4 GPa, followed by rapid ascent to the surface, but do not necessarily require high temperatures or high degrees of partial melting. Tectonically, these picrites formed in thick Archean and Paleoproterozoic cratonic terranes during Paleogene rifting between Greenland and North America. Structurally, the picrites are related to the major intersection of a NNW suture zone under Baffin Bay and the E-W trending Paleoproterozoic Nagssugtoqidian Fold Belt. During the late Mesozoic, ENE extension created normal faulted basins quasi-parallel with the NNW suture and thinned the mantle lithosphere. Elastic finite element models and present day studies of crustal extension show that the thicker Nagssugtoqidian Fold Belt underwent less thinning and extension than the NNW suture zone in the Archean Rae craton. These extensional disparities occur at the orthogonal intersection of pre-existing E-W trending strike-slip faults in the thicker Nagssugtoqidian Fold Belt with the NNW thinned Archean suture zone, and likely resulted in the formation of one or more pull-apart basins. Because the strike-slip faults are ancient suture zones, trans-tension within these suture zones easily reached 120 km, creating not only decompression melting in the SCLM, but also a pathway for the picritic melts to rapidly reach the surface. Such a purely tectonic model requires no spatially or temporally improbable deep mantle plume for generation of

  11. Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1)

    Energy Technology Data Exchange (ETDEWEB)

    Jonah, S.A. [Reactor Engineering Section, Centre for Energy Research and Training, Ahmadu Bello University, Zaria, P.M.B. 1014 (Nigeria)], E-mail: jonahsa2001@yahoo.com; Liaw, J.R.; Matos, J.E. [RERTR Program, Nuclear Engineering Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States)

    2007-12-15

    The Monte Carlo N-Particle (MCNP) code, version 4C (MCNP4C) and a set of neutron cross-section data were used to develop an accurate three-dimensional computational model of the Nigeria Research Reactor-1 (NIRR-1). The geometry of the reactor core was modeled as closely as possible including the details of all the fuel elements, reactivity regulators, the control rod, all irradiation channels, and Be reflectors. The following reactor core physics parameters were calculated for the present highly enriched uranium (HEU) core: clean cold core excess reactivity ({rho}{sub ex}), control rod (CR) and shim worth, shut down margin (SDM), neutron flux distributions in the irradiation channels, reactivity feedback coefficients and the kinetics parameters. The HEU input model was validated by experimental data from the final safety analyses report (SAR). The model predicted various key neutronics parameters fairly accurately and the calculated thermal neutron fluxes in the irradiation channels agree with the values obtained by foil activation method. Results indicate that the established Monte Carlo model is an accurate representation of the NIRR-1 HEU core and will be used to perform feasibility for conversion to low enriched uranium (LEU)

  12. Safety from the operator's perspective: We are all in this together

    International Nuclear Information System (INIS)

    Ellis, J.

    2005-01-01

    Following the Three Mile Island accident, the U.S. nuclear industry recognized that all nuclear utilities are affected by the performance of any one utility - that they are hostages of each other. This led to the formation of INPO, a unique model of self-regulation through peer review. As part of the industry's pursuit of excellence, INPO promotes a strong safety culture at each member utility. Nuclear stations need a strong safety culture because of the unique nature of the technology - the presence of radioactive byproducts and decay heat, and the concentration of energy in the reactor core. INPO's evaluation program is an intentionally intrusive process that provides comprehensive insight about a nuclear station's safety culture. The foundation for the program is the 'Performance Objectives and Criteria', which contains standards for plant and corporate performance. It is a behavior-based 'safety checklist' that INPO evaluators use in the field as they observe people at work in the plant, in the control room, during training, and in meetings. Open, candid discussions about safety culture are held with the plant staff, senior utility management, and within INPO. The 2002 discovery of degradation of the Davis-Besse Nuclear Power Station reactor vessel head highlighted problems that result when the safety environment at a plant receives insufficient attention. It also served as a stark reminder that safety culture is perishable and must constantly be rebuilt. As a result, INPO has improved its ability to detect declining plant performance, which will help the industry prevent safety significant events in the future. Promoting and evaluating safety culture has always been fundamental to INPO's work. While it has been called different things over the years (operational excellence, professionalism, conservative decision-making, or reactivity management), ensuring that nuclear safety has the overriding priority is woven into the fabric of all INPO activities. (author)

  13. Welding of the A1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Becka, J.

    1975-01-01

    As concerns welding, the A-1 reactor pressure vessel represents a geometrically complex unit containing 1492 welded joints. The length of welded sections varies between 10 and 620 mm. At an operating temperature of 120 degC and a pressure of 650 N/cm 2 the welded joints in the reactor core are exposed to an integral dose of 3x10 18 n/cm 2 . The chemical composition is shown for pressure vessel steel as specified by CSN 413090.9 modified by Ni, Ti and Al additions, and for the welding electrodes used. The requirements are also shown for the mechanical properties of the base and the weld metals. The technique and conditions of welding are described. No defects were found in ultrasonic testing of welded joints. (J.B.)

  14. Determination flux in the Reactor JEN-1

    International Nuclear Information System (INIS)

    Manas Diaz, L.; Montes Ponce de leon, J.

    1960-01-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

  15. Comparison of the N Reactor and Ignalina Unit No. 2 Level 1 Probabilistic Safety Assessments

    International Nuclear Information System (INIS)

    Coles, G.A.; McKay, S.L.

    1995-06-01

    A multilateral team recently completed a full-scope Level 1 Probabilistic Safety Assessment (PSA) on the Ignalina Unit No. 2 reactor plant in Lithuania. This allows comparison of results to those of the PSA for the U.S. Department of Energy's (DOE) N Reactor. The N Reactor, although unique as a Western design, has similarities to Eastern European and Soviet graphite block reactors

  16. Study of dietary supplements compositions by neutron activation analysis at the VR-1 training reactor

    Science.gov (United States)

    Stefanik, Milan; Rataj, Jan; Huml, Ondrej; Sklenka, Lubomir

    2017-11-01

    The VR-1 training reactor operated by the Czech Technical University in Prague is utilized mainly for education of students and training of various reactor staff; however, R&D is also carried out at the reactor. The experimental instrumentation of the reactor can be used for the irradiation experiments and neutron activation analysis. In this paper, the neutron activation analysis (NAA) is used for a study of dietary supplements containing the zinc (one of the essential trace elements for the human body). This analysis includes the dietary supplement pills of different brands; each brand is represented by several different batches of pills. All pills were irradiated together with the standard activation etalons in the vertical channel of the VR-1 reactor at the nominal power (80 W). Activated samples were investigated by the nuclear gamma-ray spectrometry technique employing the semiconductor HPGe detector. From resulting saturated activities, the amount of mineral element (Zn) in the pills was determined using the comparative NAA method. The results show clearly that the VR-1 training reactor is utilizable for neutron activation analysis experiments.

  17. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    West, C D [comp.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN{sub 2} test, Source LH2-H{sub 2}O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface.

  18. IGORR 1: Proceedings of the 1. meeting of the International Group On Research Reactors

    International Nuclear Information System (INIS)

    West, C.D.

    1990-05-01

    Descriptions of the ongoing projects presented at this Meeting were concerned with: New Research Reactor FRM-II at Munich; MITR-II reactor; The Advanced. Neutron Source (ANS) Project; The high Flux Reactor Petten, Status and Prospects; The High Flux Beam Reactor Instrumentation Upgrade; BER-II Upgrade; The BR2 Materials Testing Reactor Past, Ongoing and Under-Study Upgrades; The ORPHEE, Reactor Current Status and Proposed Enhancement of Experimental Variabilities; Construction of the Upgraded JRR-3; Status of the University of Missouri-Columbia Research Reactor Upgrade; the Reactor and Cold Neutron Facility at NIST; Upgrade of Materials Irradiation Facilities in HFIR; Backfitting of the FRG Reactors; University Research Reactors in the United States; and Organization of the ITER Project - Sharing of Informational Procurements. Topics of interest were: Thermal-hydraulic tests and correlations, Corrosion tests and analytical models , Multidimensional kinetic analysis for small cores, Fuel plates fabrication, Fuel plates stability, Fuel irradiation, Burnable poison irradiation, Structural materials irradiation, Neutron guides irradiation, Cold Source materials irradiation, Cold Source LN 2 test, Source LH2-H 2 O reaction (H or D), Instrumentation upgrading and digital control system, Man-machine interface

  19. Nuclear reactor and materials science research: Technical report, May 1, 1985-September 30, 1986

    International Nuclear Information System (INIS)

    1987-01-01

    Throughout the 17-month period of its grant, May 1, 1985-September 30, 1986, the MIT Research Reactor (MITR-II) was operated in support of research and academic programs in the physical and life sciences and in related engineering fields. The reactor was operated 4115 hours during FY 1986 and for 6080 hours during the entire 17-month period, an average of 82 hours per week. Utilization of the reactor during that period may be classified as follows: neutron beam tube research; nuclear materials research and development; radiochemistry and trace analysis; nuclear medicine; radiation health physics; computer control of reactors; dose reduction in nuclear power reactors; reactor irradiations and services for groups outside MIT; MIT Research Reactor. Data on the above utilization for FY 1986 show that the MIT Nuclear Reactor Laboratory (NRL) engaged in joint activities with nine academic departments and interdepartmental laboratories at MIT, the Charles Stark Draper Laboratory in Cambridge, and 22 other universities and nonprofit research institutions, such as teaching hospitals

  20. Baikal-1 stand complex. Preparation and carrying out of the first energy start-up of the IVG-1 reactor

    International Nuclear Information System (INIS)

    Tikhomirov, L.N.

    1995-01-01

    The IVG-1 reactor was a first ground prototype of nuclear rocket engine. The reactor was built on the site 10 of the Semipalatinsk test site. Since the first energy start-up in 1975 the reactor was exploited 14 years till its modernization in 1989. The Bajkal-1 stand complex was designed and built for the carrying out of tests for fuel assemblies of different modifications. The energy start-up has been sum of long creative work of different research and constructive staffs on creation of high-temperature gas-cooled IVG-1 reactor. The history of construction, project and assembling of the stand complex is presented. Complex start and put works were carried out in the December 1974. Control physical start-up was carried out in the January 1975. Cold start-up by hydrogen was in the February 1975. Hot start-up was in the March 1975. The result of the hot start-up was experimental confirmation of metodics of thermohydrovlical estimations. 2 figs., 3 tabs

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  2. Empirical and Computational Support for Context-Dependent Representations of Serial Order: Reply to Bowers, Damian, and Davis (2009)

    Science.gov (United States)

    Botvinick, Matthew M.; Plaut, David C.

    2009-01-01

    J. S. Bowers, M. F. Damian, and C. J. Davis (2009) critiqued the computational model of serial order memory put forth in M. Botvinick and D. C. Plaut (2006), purporting to show that the model does not generalize in a way that people do. They attributed this supposed failure to the model's dependence on context-dependent representations,…

  3. Determination flux in the Reactor JEN-1; Medida de flujos de neutrones en el nucleo del Reactor JEN-1

    Energy Technology Data Exchange (ETDEWEB)

    Manas Diaz, L; Montes Ponce de leon, J.

    1960-07-01

    This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 {mu} gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs.

  4. Reactor core conversion studies of Ghana: Research Reactor-1 and proposal for addition of safety rod

    International Nuclear Information System (INIS)

    Odoi, H.C.

    2014-06-01

    The inclusion of an additional safety rod in conjunction with a core conversion study of Ghana Research Reactor-1 (GHARR-1) was carried out using neutronics, thermal hydraulics and burnup codes. The study is based on a recommendation by Integrated Safety Assessment for Research Reactors (INSARP) mission to incorporate a safety rod to the reactor safety system as well as the need to replace the reactor fuel with LEU. Conversion from one fuel type to another requires a complete re-evaluation of the safety analysis. Changes to the reactivity worth, shutdown margin, power density and material properties must be taken into account, and appropriate modifications made. Neutronics analysis including burnup was studied followed by thermal hydraulics analyses which comprise steady state and transients. Four computer codes were used for the analysis; MCNP, REBUS, PLTEP and PARET. The neutronics analysis revealed that the LEU core must be operated at 34 Kw in order to attain the flux of 1.0E12 n/cm 2 .s as the nominal flux of the HEU core. The auxiliary safety rod placed at a modified irradiation site gives a better worth than the cadmium capsules. For core excess reactivity of 4 mk, 348 fuel pins would be appropriate for the GHARR-1 LEU core. Results indicate that flux level of 1.0E12 n/cm 2 .s in the inner irradiation channel will not be compromised, if the power of the LEU core is increased to 34 kW. The GHARR-1 core using LEU-U0 2 -12.5% fuel can be operated for 23 shim cycles, with cycles length 2.5 years, for over 57 years at the 17 kW power level. All 23 LEU cycles meet the ∼ 4.0 mk excess reactivity required at the beginning of cycle . For comparison, the MNSR HEU reference core can also be operated for 23 shim cycles, but with a cycle length of 2.0 years for just over 46 years at 15.0kW power level. It is observed that the GHARR-1 core with LEU UO 2 fuel enriched to 12.5% and a power level of 34 kW can be operated ∼25% longer than the current HEU core operated at

  5. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  6. Reactor Engineering Department annual report, April 1, 1985 - March 31, 1986

    International Nuclear Information System (INIS)

    1986-08-01

    Research and development activities in the Department of Reactor Engineering in fiscal 1985 are described. The work of the Department is closely related to development of multipurpose Very High Temperature Gas Cooled Reactor, High Conversion Light Water 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, reactor decommissioning technology, and activities of the Committee on Reactor Physics. (author)

  7. Economics and utilization of thorium in nuclear reactors. Technical annexes 1 and 2

    International Nuclear Information System (INIS)

    1978-05-01

    An assessment of the impact of utilizing the 233 U/thorium fuel cycle in the U.S. nuclear economy is strongly dependent upon several decisions involving nuclear energy policy. These decisions include: (1) to recycle or not recycle fissile material; (2) if fissile material is recycled, to recycle plutonium, 233 U, or both; and (3) to deploy or not to deploy advanced reactor designs such as Fast Breeder Reactors (FBR's), High Temperature Gas Reactors (HTGR's), and Canadian Deuterium Uranium Reactors (CANDU's). This report examines the role of thorium in the context of the above policy decisions while focusing special attention on economics and resource utilization

  8. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  9. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Shal'nov, A.V.

    1995-01-01

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  10. Experiment on continuous operation of the Brazilian IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Freitas Pintaud, M. de

    1994-01-01

    In order to increase the radioisotope production in the IEA-R1 research reactor at IPEN/CNEN-SP, it has been proposed a change in its operation regime from 8 hours per day and 5 days per week to continuous 48 hours per week. The necessary reactor parameters for this new operation regime were obtained through an experiment in which the reactor was for the first time operated in the new regime. This work presents the principal results from this experiment: xenon reactivity, new shutdown margins, and reactivity loss due to fuel burnup in the new operation regime. (author)

  11. Auxiliary control system of the safety parameters for IPR-R1 reactor

    International Nuclear Information System (INIS)

    Coura, J.G.

    1986-01-01

    This paper deals with the description for the control of three cooling water parameters (conductivity, temperature and the maximum and minimum water levels) as well as the percent power fraction of the nuclear research reactor IPR-R1. In order to keep the reactor in good operation conditions, one permanent and accurate control of the cooling water is needed. The double monitoring of a fourth parameter, part of the original design, the percent power fraction, is obtained through the control of the uncompensated ion chamber current and aims to avoid the operation of the reactor without running the cooling system. (Author) [pt

  12. OSCAR-4 Code System Application to the SAFARI-1 Reactor

    International Nuclear Information System (INIS)

    Stander, Gerhardt; Prinsloo, Rian H.; Tomasevic, Djordje I.; Mueller, Erwin

    2008-01-01

    The OSCAR reactor calculation code system consists of a two-dimensional lattice code, the three-dimensional nodal core simulator code MGRAC and related service codes. The major difference between the new version of the OSCAR system, OSCAR-4, and its predecessor, OSCAR-3, is the new version of MGRAC which contains many new features and model enhancements. In this work some of the major improvements in the nodal diffusion solution method, history tracking, nuclide transmutation and cross section models are described. As part of the validation process of the OSCAR-4 code system (specifically the new MGRAC version), some of the new models are tested by comparing computational results to SAFARI-1 reactor plant data for a number of operational cycles and for varying applications. A specific application of the new features allows correct modeling of, amongst others, the movement of fuel-follower type control rods and dynamic in-core irradiation schedules. It is found that the effect of the improved control rod model, applied over multiple cycles of the SAFARI-1 reactor operation history, has a significant effect on in-cycle reactivity prediction and fuel depletion. (authors)

  13. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    International Nuclear Information System (INIS)

    Novelli, A.

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer. (author)

  14. L’enseignement clinique du droit de l’immigration aux Etats-Unis : la clinique de l’Université de Californie Davis et le programme DACA

    OpenAIRE

    Kahssay, Jihan A.

    2014-01-01

    Les cliniques de l'immigration jouent un rôle important et original dans l'enseignement du droit de l'immigration aux USA. Elle remplissent également un rôle social important dans l'accès au droit pour les publics défavorisés. Dans cette interview, J. Kahssay, avocate et chargée du programme DACA à la clinique du droit de l'immigration de l'Université de Californie Davis (King hall school of law), présente le fonctionnement général de la clinique du droit de l'immigration de UC Davis. Elle dé...

  15. Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

    International Nuclear Information System (INIS)

    Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

    2008-01-01

    The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

  16. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  17. SCALE-4 Analysis of LaSalle Unit 1 BWR Commercial Reactor Critical Configurations

    International Nuclear Information System (INIS)

    Gauld, I.C.

    2000-01-01

    Five commercial reactor criticals (CRCs) for the LaSalle Unit 1 boiling-water reactor have been analyzed using KENO V.a, the Monte Carlo criticality code of the SCALE 4 code system. The irradiated fuel assembly isotopics for the criticality analyses were provided by the Waste Package Design team at the Yucca Mountain Project in the US, who performed the depletion calculations using the SAS2H sequence of SCALE 4. The reactor critical measurements involved two beginning-of-cycle and three middle-of-cycle configurations. The CRCs involved relatively low-cycle burnups, and therefore contained a relatively high gadolinium poison content in the reactor assemblies. This report summarizes the data and methods used in analyzing the critical configurations and assesses the sensitivity of the results to some of the modeling approximations used to represent the gadolinium poison distribution within the assemblies. The KENO V.a calculations, performed using the SCALE 44GROUPNDF5 ENDF/B-V cross-section library, yield predicted k eff values within about 1% Δk/k relative to reactor measurements for the five CRCs using general 8-pin and 9-pin heterogeneous gadolinium poison pin assembly models

  18. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor; Calculo de vairaciones de reactividad en algunos periodos regulares de operacion del reactor JEN-1 Mod.

    Energy Technology Data Exchange (ETDEWEB)

    Alcala Ruiz, F

    1973-07-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  19. Reactor Engineering Department annual report (April 1, 1986 - March 31, 1987)

    International Nuclear Information System (INIS)

    1987-08-01

    Research and development activities in the Department of Reactor Engineering in the fiscal year 1986 are described. The major activities of the Department are closely related to the reactor physics of very high temperature gas-cooled reactor, high conversion light water reactor and liquid metal fast breeder reactor and to blanket neutronics of fusion reactor. Contents of this report are divided into the activities on nuclear data and group constants, theoretical methods and code development, reactor physics experiment and analysis, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control, diagnosis and robotics. The activity of the Research Committee on Reactor Physics is also included. (author)

  20. 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.)

  1. Reactor Physics Programme

    Energy Technology Data Exchange (ETDEWEB)

    De Raedt, C

    2000-07-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

  2. Reactor Physics Programme

    International Nuclear Information System (INIS)

    De Raedt, C.

    2000-01-01

    The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

  3. Modelling of the RA-1 reactor using a Monte Carlo code; Modelado del reactor RA-1 utilizando un codigo Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Quinteiro, Guillermo F; Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Reactores y Centrales Nucleares

    2000-07-01

    It was carried out for the first time, a model of the Argentine RA-1 reactor using the MCNP Monte Carlo code. This model was validated using data for experimental neutron and gamma measurements at different energy ranges and locations. In addition, the resulting fluxes were compared with the data obtained using a 3D diffusion code. (author)

  4. Neutronics and thermohydraulics of the reactor C.E.N.E. Pt. 1

    International Nuclear Information System (INIS)

    Caro, R.; Ahnert, C.; Esteban Naudin, A.; Martinez Fanegas, R.; Minguez, E.; Rovira, A.

    1976-01-01

    The analysis of neutronics (both statics and kinetics), of the 10 Mwt swimming pool reactor C.E.N.E. is included. A short description of the theoretical model used, along with the theoretical versus experimental cheking, carried out, whenever possible, with the reactors JEN-1 and JEN-2 of Junta de Energia Nuclear, is given in each of these chapters. (author) [es

  5. Applied research into direct numerical control of A-1 reactor temperature

    International Nuclear Information System (INIS)

    Karpeta, C.; Volf, K.

    1974-01-01

    Partial results of research efforts aimed at applying modern control theory in the control of the reactor of the A-1 nuclear power station are presented. A mathematical model of the process dynamics was developed. Some parameters of the model were determined using the results of an experimentally performed reactor scram. The optimal stochastic discrete regulator was determined and closed-loop transients were studied. The possibilities of implementing control routines were investigated using the RPP-16 computer. (author)

  6. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

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

    International Nuclear Information System (INIS)

    1990-01-01

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

  8. Research reactor core conversion guidebook. V.1: Summary

    International Nuclear Information System (INIS)

    1992-04-01

    In view of the proliferation concerns caused by the use of highly enriched uranium (HEU) and in anticipation that the supply of HEU to research and test reactors will be more restricted in the future, this guidebook has been prepared to assist research reactor operators in addressing the safety and licensing issues for conversion of their reactor cores from the use of HEU fuel to the use of low enriched uranium fuel. This Guidebook, in five volumes, addresses the effects of changes in the safety-related parameters of mixed cores and the converted core. It provides an information base which should enable the appropriate approvals processes for implementation of a specific conversion proposal, whether for a light or for a heavy water moderated research reactor. Refs, figs, bibliographies and tabs

  9. Satisfying the diverse development needs of an engineering organization

    International Nuclear Information System (INIS)

    Zarkesh, L.P.

    1991-01-01

    The Engineering Department at Davis-Besse Nuclear Power Station has established an aggressive philosophy for professionally developing their staff. This philosophy has evolved over the last four years into a program with specialized administrative tools which not only satisfies the intent of industry training guidelines, but also accentuates the development of the individual. This program consists of three parts: (1) The Development Program - A program constructed to actively integrate system and applied science courses, management and interpersonal skill courses, design basis courses (e.g., pipe break analysis, support design, etc.) special process courses (e.g., human factors, ALARA, etc.) and external seminars sponsored by industry experts; (2) The Individual Development Plan (IDP) - A documented course of action, developed annually, in which the employee and the line supervisor jointly contribute to the identification of career goals and strategic professional objectives; and (3) The Training Database - A PC database developed to retain and manage course information (e.g., requests, attendance priorities, schedule, history, etc.). The paper describes these three facets of the training program

  10. 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

  11. 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.

  12. NBR ISO 9001 Certification for activities carried out in IEA-R1 reactor

    International Nuclear Information System (INIS)

    Paiva, Rosemeire P.; Salvetti, Tereza C.

    2005-01-01

    Since its inauguration in 1957, the IEA-R1 research reactor has been used mainly for research, development and teaching by scientific community. In the last years, with the increase of the commercial radiopharmaceutical production by Radiopharmacy Center of IPEN, the IEA-R1 reactor was recognized as a service supplier for that center and has received a treatment more commercial from IPEN Management. In 1999 the radiopharmaceutical production obtained the NBR ISO 9002 Certification, since that the IPEN Management considered convenient to invest in the certification of its internal suppliers. In this context, in 2001 the Research Reactor Center (CRPq) began the implantation of a Quality Management System (QMS) based on NBR 9001: 2000 standard, for activities related to the operation and maintenance of the IEA-R1 research reactor and irradiation services. This QMS was structured to incorporate tools already implemented in order to complain the requirements related to nuclear and radiological safe for a nuclear installation established by the regulatory organism. The QMS is supported by a documentation system composed of approximately 150 documents including quality manual, business and action plans, operational procedures and work instruction. Carlos Alberto Vanzolini Foundation (FCAV), an INMETRO certified organism, certified the 'Operation and Maintenance of the IEA-R1 Research Reactor and Irradiation Services' in December 2002. In 2003 and 2004, the QMS was audited by FCAV that determined the maintenance of the certification. This work presents the main steps of the QMS implementation, including the difficulties found and results obtained in the process. (author)

  13. Field tests experience from 1.6MW/400kWh Li-ion battery energy storage system providing primary frequency regulation service

    DEFF Research Database (Denmark)

    Swierczynski, Maciej Jozef; Stroe, Daniel Ioan; Stan, Ana-Irina

    2013-01-01

    Lithium-ion battery energy storage systems (BESSs) represent suitable alternatives to conventional generating units for providing primary frequency regulation on the Danish market. This paper presents aspects concerning the operation of the BESSs in the Danish energy market while providing upwards...... on the BESS demonstrator located in Western Denmark and initial results are introduced and discussed. These measurements can be used to validate models for battery ageing during realistic operation or to develop the diagnostic tools for the BESS....

  14. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    Rowan, G.A.; Sun, J.B.; Mott, S.L.

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  15. "Our sickness record is a national disgrace": Adelle Davis, nutritional determinism, and the anxious 1970s.

    Science.gov (United States)

    Carstairs, Catherine

    2014-07-01

    America's most widely read nutritionist of the postwar decades, Adelle Davis, helped to shape Americans' eating habits, their child-feeding practices, their views about the quality of their food supply, and their beliefs about the impact of nutrition on their emotional and physical health. This paper closely examines Davis's writings and argues that even though she is often associated with countercultural food reformers like Alice Waters and Frances Moore Lappé, she had as much in common with the writings of interwar nutritionists and home economists. While she was alarmed about the impact of pesticides and food additives on the quality of the food supply, and concerned about the declining fertility of American soil, she commanded American women to feed their families better and promised that improved nutrition would produce stronger, healthier, more beautiful children who would ensure America's future strength. She believed that nearly every health problem could be solved through nutrition, and urged her readers to manage their diets carefully and to take extensive supplementation to ensure optimum health. As such, she played an important role in creating the ideology of "nutritionism" - the idea that food should be valued more for its constituent parts than for its pleasures or cultural significance. © The Author 2012. Published by Oxford University Press. All rights reserved. For permissions, please e-mail: journals.permissions@oup.com.

  16. Reactor calculations in aid of isotope production at SAFARI-1

    International Nuclear Information System (INIS)

    Ball, G.

    2003-01-01

    Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

  17. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  18. Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

    International Nuclear Information System (INIS)

    Chambadal, P.; Pascal, M.

    1955-01-01

    The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr

  19. Development of a training simulator to operators of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Carvalho, Ricardo Pinto de

    2006-01-01

    This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

  20. An overview of the RECH-1 reactor conversion

    International Nuclear Information System (INIS)

    Klein, J.; Medel, J.; Daie, J.; Torres, H.

    2000-01-01

    The RECH-l research reactor achieved the first criticality on October 13, 1974 using HEU MTR type fuel elements, which were fabricated by the UKAEA at Dounreay, Scotland. In 1979, the conversion of the reactor to use LEU fuel was decided; however, a rough estimate of the uranium density needed to convert the reactor gave 3.7 g/cm 3 . This density was not available, and to maintain the overall fuel element geometry it was necessary to convert the reactor to use 45% enriched uranium fuel. In 1985, the conversion of the reactor to use medium enriched uranium was achieved. Some years later, the Chilean Nuclear Energy Commission developed the capability to produce fuel elements based on U 3 Si 2 -Al dispersion fuel. Once the plant and the manufacturing and quality control procedures were commissioned to permit the production of fuel elements, a fabrication program starts to produce LEU fuel elements with a uranium density of 3.4 g/cm 3 . A fabrication qualification period that extended to the required fuel plates for the assembly of two fuel elements started. In November 1998, the first four LEU fuel elements manufactured by the Chilean Fuel Fabrication Plant were delivered to the reactor. When the first two fuel elements were introduced into the core a LEU fuel element qualification program began. While those fuel elements remain in the core, an evaluation program is being applied to observe its performance under irradiation condition. (author)

  1. Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

    International Nuclear Information System (INIS)

    Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

    2005-01-01

    G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

  2. Verification of the linearity of the IPR-R1 TRIGA reactor power channels

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

    2013-01-01

    The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

  3. Conceptual design study for the demonstration reactor of JSFR. (1) Current status of JSFR development

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Sakamoto, Yoshihiko; Kotake, Shoji; Aoto, Kazumi; Ohshima, Jun; Ito, Takaya

    2011-01-01

    JAEA is now conducting 'Fast Reactor Cycle Technology Development (FaCT)' project for the commercialization before 2050s. A demonstration reactor of Japan Sodium-cooled Fast Reactor (JSFR) is planned to start operation around 2025. In the FaCT project, conceptual design study on the demonstration reactor has been performed since 2007 to determine the referential reactor specifications for the next stage design work from 2011 for the licensing and construction. Plant performance as a demonstration reactor for the 1.5 GWe commercial reactor JSFR is being compared between 750 MWe and 500 MWe plant designs. By using the results of conceptual design study, output power will be determined during year of 2010. This paper describes development status of key technologies and comparison between 750 MWe and 500 MWe plants with the view points of demonstration ability for commercial JSFR plant. (author)

  4. An economic analysis of stretch-out for Angra-1 reactor

    International Nuclear Information System (INIS)

    Sakai, M.

    1989-01-01

    An application of NUCOST code for calculating nuclear energy cost is presented. Ann optimization of stretch-out for Angra-1 reactor based on international costs of nuclear fuel, operation and maintenance is done. (M.C.K.)

  5. The analysis for inventory of experimental reactor high temperature gas reactor type

    International Nuclear Information System (INIS)

    Sri Kuntjoro; Pande Made Udiyani

    2016-01-01

    Relating to the plan of the National Nuclear Energy Agency (BATAN) to operate an experimental reactor of High Temperature Gas Reactors type (RGTT), it is necessary to reactor safety analysis, especially with regard to environmental issues. Analysis of the distribution of radionuclides from the reactor into the environment in normal or abnormal operating conditions starting with the estimated reactor inventory based on the type, power, and operation of the reactor. The purpose of research is to analyze inventory terrace for Experimental Power Reactor design (RDE) high temperature gas reactor type power 10 MWt, 20 MWt and 30 MWt. Analyses were performed using ORIGEN2 computer code with high temperatures cross-section library. Calculation begins with making modifications to some parameter of cross-section library based on the core average temperature of 570 °C and continued with calculations of reactor inventory due to RDE 10 MWt reactor power. The main parameters of the reactor 10 MWt RDE used in the calculation of the main parameters of the reactor similar to the HTR-10 reactor. After the reactor inventory 10 MWt RDE obtained, a comparison with the results of previous researchers. Based upon the suitability of the results, it make the design for the reactor RDE 20MWEt and 30 MWt to obtain the main parameters of the reactor in the form of the amount of fuel in the pebble bed reactor core, height and diameter of the terrace. Based on the main parameter or reactor obtained perform of calculation to get reactor inventory for RDE 20 MWT and 30 MWT with the same methods as the method of the RDE 10 MWt calculation. The results obtained are the largest inventory of reactor RDE 10 MWt, 20 MWt and 30 MWt sequentially are to Kr group are about 1,00E+15 Bq, 1,20E+16 Bq, 1,70E+16 Bq, for group I are 6,50E+16 Bq, 1,20E+17 Bq, 1,60E+17 Bq and for groups Cs are 2,20E+16 Bq, 2,40E+16 Bq, 2,60E+16 Bq. Reactor inventory will then be used to calculate the reactor source term and it

  6. Welding electrode for peripheral welds of A-1 reactor pressure vessel

    International Nuclear Information System (INIS)

    Lakatos, L.

    1975-01-01

    The properties are outlined of the VUZ-AC1-52 welding electrode used in welding the Bohunice A-1 reactor pressure vessel. The mechanical properties of welded joints after the final thermal treatment are summed up. (J.K.)

  7. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  8. The reactor kinetics code tank: a validation against selected SPERT-1b experiments

    International Nuclear Information System (INIS)

    Ellis, R.J.

    1990-01-01

    The two-dimensional space-time analysis code TANK is being developed for the simulation of transient behaviour in the MAPLE class of research reactors. MAPLE research reactor cores are compact, light-water-cooled and -moderated, with a high degree of forced subcooling. The SPERT-1B(24/32) reactor core had many similarities to MAPLE-X10, and the results of the SPERT transient experiments are well documented. As a validation of TANK, a series of simulations of certain SPERT reactor transients was undertaken. Special features were added to the TANK code to model reactors with plate-type fuel and to allow for the simulation of rapid void production. The results of a series of super-prompt-critical reactivity step-insertion transient simulations are presented. The selected SPERT transients were all initiated from low power, at ambient temperatures, and with negligible coolant flow. Th results of the TANK simulations are in good agreement with the trends in the experimental SPERT data

  9. Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

    Energy Technology Data Exchange (ETDEWEB)

    Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

    2003-07-01

    By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

  10. . Effects of extended shutdown on the control rod drive mechanism of nigeria research reactor-1(NIRR-1)

    International Nuclear Information System (INIS)

    Yusuf, I; Mati, A. A.

    2010-01-01

    The control rod drive mechanism of the Nigeria Research Reactor-1 is being driven by a servo motor, type SDE-45 through a mechanical gear system. The servo motor ensures the position control of the control rod, and hence the stability of the neutron-flux of the nuclear research reactor. The control rod drive mechanism assembly is mounted on top of the reactor vessel, about 0.6m above 30m 3 volume of reactor pool water. The top of the pool is covered with a Perspex material to protect the water in the pool from environmental contamination and to reduce evaporation. Although most of the materials in the control rod drive mechanism assembly are made of stainless steel, the servo motor however contains corrodible materials. The paper reveals a practical experience of failure of the control rod drive mechanism as a result of corrosion growth between the rotor of the servo motor and its stator windings, due to an extended shutdown of the facility.

  11. Low Substrate Loading Limits Methanogenesis and Leads to High Coulombic Efficiency in Bioelectrochemical Systems

    Directory of Open Access Journals (Sweden)

    Tom H. J. A. Sleutels

    2016-01-01

    Full Text Available A crucial aspect for the application of bioelectrochemical systems (BESs as a wastewater treatment technology is the efficient oxidation of complex substrates by the bioanode, which is reflected in high Coulombic efficiency (CE. To achieve high CE, it is essential to give a competitive advantage to electrogens over methanogens. Factors that affect CE in bioanodes are, amongst others, the type of wastewater, anode potential, substrate concentration and pH. In this paper, we focus on acetate as a substrate and analyze the competition between methanogens and electrogens from a thermodynamic and kinetic point of view. We reviewed experimental data from earlier studies and propose that low substrate loading in combination with a sufficiently high anode overpotential plays a key-role in achieving high CE. Low substrate loading is a proven strategy against methanogenic activity in large-scale reactors for sulfate reduction. The combination of low substrate loading with sufficiently high overpotential is essential because it results in favorable growth kinetics of electrogens compared to methanogens. To achieve high current density in combination with low substrate concentrations, it is essential to have a high specific anode surface area. New reactor designs with these features are essential for BESs to be successful in wastewater treatment in the future.

  12. Low Substrate Loading Limits Methanogenesis and Leads to High Coulombic Efficiency in Bioelectrochemical Systems

    Science.gov (United States)

    Sleutels, Tom H. J. A.; Molenaar, Sam D.; Heijne, Annemiek Ter; Buisman, Cees J. N.

    2016-01-01

    A crucial aspect for the application of bioelectrochemical systems (BESs) as a wastewater treatment technology is the efficient oxidation of complex substrates by the bioanode, which is reflected in high Coulombic efficiency (CE). To achieve high CE, it is essential to give a competitive advantage to electrogens over methanogens. Factors that affect CE in bioanodes are, amongst others, the type of wastewater, anode potential, substrate concentration and pH. In this paper, we focus on acetate as a substrate and analyze the competition between methanogens and electrogens from a thermodynamic and kinetic point of view. We reviewed experimental data from earlier studies and propose that low substrate loading in combination with a sufficiently high anode overpotential plays a key-role in achieving high CE. Low substrate loading is a proven strategy against methanogenic activity in large-scale reactors for sulfate reduction. The combination of low substrate loading with sufficiently high overpotential is essential because it results in favorable growth kinetics of electrogens compared to methanogens. To achieve high current density in combination with low substrate concentrations, it is essential to have a high specific anode surface area. New reactor designs with these features are essential for BESs to be successful in wastewater treatment in the future. PMID:27681899

  13. Low Substrate Loading Limits Methanogenesis and Leads to High Coulombic Efficiency in Bioelectrochemical Systems.

    Science.gov (United States)

    Sleutels, Tom H J A; Molenaar, Sam D; Heijne, Annemiek Ter; Buisman, Cees J N

    2016-01-05

    A crucial aspect for the application of bioelectrochemical systems (BESs) as a wastewater treatment technology is the efficient oxidation of complex substrates by the bioanode, which is reflected in high Coulombic efficiency (CE). To achieve high CE, it is essential to give a competitive advantage to electrogens over methanogens. Factors that affect CE in bioanodes are, amongst others, the type of wastewater, anode potential, substrate concentration and pH. In this paper, we focus on acetate as a substrate and analyze the competition between methanogens and electrogens from a thermodynamic and kinetic point of view. We reviewed experimental data from earlier studies and propose that low substrate loading in combination with a sufficiently high anode overpotential plays a key-role in achieving high CE. Low substrate loading is a proven strategy against methanogenic activity in large-scale reactors for sulfate reduction. The combination of low substrate loading with sufficiently high overpotential is essential because it results in favorable growth kinetics of electrogens compared to methanogens. To achieve high current density in combination with low substrate concentrations, it is essential to have a high specific anode surface area. New reactor designs with these features are essential for BESs to be successful in wastewater treatment in the future.

  14. VR-1 training reactor in use for twelve years to train experts for the Czech nuclear power sector

    International Nuclear Information System (INIS)

    Matejka, K.; Sklenka, L.

    2003-01-01

    The VR-1 training reactor has been serving students of the Faculty of Nuclear Science and Physical Engineering, Czech Technical University in Prague, for more than 12 years now. The operation history of the reactor is highlighted. The major changes made at the VR-1 reactor are outlined and the main experimentally verified core configurations are shown. Some components of the new equipment installed on the VR-1 reactor are described in detail. The fields of application are shown: the reactor serves not only the training of university students within whole Czech Republic but also the training of specialists, research activities, and information programmes in the nuclear power domain. (P.A.)

  15. FiR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production

    International Nuclear Information System (INIS)

    Auterinen, I.; Salmenhaara, S.E.J. . Author

    2004-01-01

    The FiR 1 reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose for the existence of the reactor is now the Boron Neutron Capture Therapy (BNCT), but FiR 1 has also an important national role in providing local enterprises and research institutions in the fields of industrial measurements, pharmaceuticals, electronics etc. with isotope production and activation analysis services. In the 1990's a BNCT treatment facility was built at the FiR 1 reactor located at Technical Research Centre of Finland. A special new neutron moderator material Fluental TM (Al+AlF3+Li) developed at VTT ensures the superior quality of the neutron beam. Also the treatment environment is of world top quality after a major renovation of the whole reactor building in 1997. Recently the lithiated polyethylene neutron shielding of the beam aperture was modified to ease the positioning of the patient close to the beam aperture. Increasing the reactor power to 500 kW would allow positioning of the patient further away from the beam aperture. Possibilities to accomplish a safety analysis for this is currently under considerations. Over thirty patients have been treated at FiR 1 since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization, Boneca Corporation. Currently three clinical trial protocols for tumours in the brain as well as in the head and neck region are recruiting patients. (author)

  16. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

  17. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  18. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  19. Nuclear material control at IEA-R1 nuclear research reactor

    International Nuclear Information System (INIS)

    1988-01-01

    The control measurements system and verification of physical inventory for fuel elements used in the operation of IEA-R1 nuclear research reactor are described. The computer code used for burn-up calculation are shown. (E.G.) [pt

  20. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

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

    International Nuclear Information System (INIS)

    1997-01-01

    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: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. 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 Associate 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 1% 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

  2. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    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: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. 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 Associate 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 1% 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.

  3. Reliability database of IEA-R1 Brazilian research reactor: Applications to the improvement of installation safety

    International Nuclear Information System (INIS)

    Oliveira, P.S.P.; Tondin, J.B.M.; Martins, M.O.; Yovanovich, M.; Ricci Filho, W.

    2010-01-01

    In this paper the main features of the reliability database being developed at Ipen-Cnen/SP for IEA-R1 reactor are briefly described. Besides that, the process for collection and updating of data regarding operation, failure and maintenance of IEA-R1 reactor components is presented. These activities have been conducted by the reactor personnel under the supervision of specialists in Probabilistic Safety Analysis (PSA). The compilation of data and subsequent calculation are based on the procedures defined during an IAEA Coordinated Research Project which Brazil took part in the period from 2001 to 2004. In addition to component reliability data, the database stores data on accident initiating events and human errors. Furthermore, this work discusses the experience acquired through the development of the reliability database covering aspects like improvements in the reactor records as well as the application of the results to the optimization of operation and maintenance procedures and to the PSA carried out for IEA-R1 reactor. (author)

  4. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

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

    2010-01-01

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

  5. 1DB, a one-dimensional diffusion code for nuclear reactor analysis

    International Nuclear Information System (INIS)

    Little, W.W. Jr.

    1991-09-01

    1DB is a multipurpose, one-dimensional (plane, cylinder, sphere) diffusion theory code for use in reactor analysis. The code is designed to do the following: To compute k eff and perform criticality searches on time absorption, reactor composition, reactor dimensions, and buckling by means of either a flux or an adjoint model; to compute collapsed microscopic and macroscopic cross sections averaged over the spectrum in any specified zone; to compute resonance-shielded cross sections using data in the shielding factor formnd to compute isotopic burnup using decay chains specified by the user. All programming is in FORTRAN. Because variable dimensioning is employed, no simple restrictions on problem complexity can be stated. The number of spatial mesh points, energy groups, upscattering terms, etc. is limited only by the available memory. The source file contains about 3000 cards. 4 refs

  6. Comparison of applicability of current transition temperature shift models to SA533B-1 reactor pressure vessel steel of Korean nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Ji Hyun; Lee, Bong Sang [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-08-15

    The precise prediction of radiation embrittlement of aged reactor pressure vessels (RPVs) is a prerequisite for the long-term operation of nuclear power plants beyond their original design life. The expiration of the operation licenses for Korean reactors the RPVs of which are made from SA533B-1 plates and welds is imminent. Korean regulatory rules have adopted the US Nuclear Regulatory Commission's transition temperature shift (TTS) models to the prediction of the embrittlement of Korean reactor pressure vessels. The applicability of the TTS model to predict the embrittlement of Korean RPVs made of SA533B-1 plates and welds was investigated in this study. It was concluded that the TTS model of 10 CFR 50.61a matched the trends of the radiation embrittlement in the SA533B-1 plates and welds better than did that of Regulatory Guide (RG) 1.99 Rev. 2. This is attributed to the fact that the prediction performance of 10 CFR 50.61a was enhanced by considering the difference in radiation embrittlement sensitivity among the different types of RPV materials.

  7. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  8. The progressive achievement of a closed fuel cycle in France; La mise en oeuvre progressive d'un cycle ferme en France

    Energy Technology Data Exchange (ETDEWEB)

    Hugelmann, D.; Devezeaux de Lavergne, J.G. [AREVA NC, 78 - Velizy Villacoublay (France)

    2008-03-15

    The author reviews the progressive building of a strong nuclear fuel cycle industry in France. The first major step was the abandon of the graphite-gas reactor system to the PWR system. The government's decision to opt for reactors operating with enriched uranium opened the way to the application at an industrial scale of uranium enrichment technology that was only confined to military purposes. The legal entity 'EURODIF S A' was founded at that time and the different production units of the George-Besse-1 enrichment plant entered into service progressively from 1978 to 1982. The Comurhex company was created in 1969, and was in charge of producing the uranium hexafluoride necessary to the fabrication of nuclear fuels. La-Hague plant entered into service in 1966, its aim was to process spent fuels from graphite-gas reactors. Inside this plant the HAO (High Activity Oxide) dedicated to PWR spent fuels was operating in 1974. The MELOX plant dedicated to the fabrication of mixed oxides fuels (Mox) entered into operation in 1995 (till now more than 5000 Mox assemblies have been fabricated. Another important step was the processing of Mox fuels. During these 30 years, the nuclear industry has made impressing progress concerning: the increase of burn-up rates, the performance of fuels, the increase in the volume being processed, the packaging of radioactive wastes, the development of nuclear transport, and a reduction of the impact on the environment. In order to maintain its level of performance the nuclear industry has made important investments concerning: mining (a global investment of 2.3*10{sup 9} euros), Comurhex-2 (a 610*10{sup 6} euros investment) and Georges-Besse-2 plant (a 3*10{sup 9} euros investment for the enrichment of uranium through centrifugation). (A.C.)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  10. Reactor Physics Training

    International Nuclear Information System (INIS)

    Baeten, P.

    2007-01-01

    University courses in nuclear reactor physics at the universities consist of a theoretical description of the physics and technology of nuclear reactors. In order to demonstrate the basic concepts in reactor physics, training exercises in nuclear reactor installations are also desirable. Since the number of reactor facilities is however strongly decreasing in Europe, it becomes difficult to offer to students a means for demonstrating the basic concepts in reactor physics by performing training exercises in nuclear installations. Universities do not generally possess the capabilities for performing training exercises. Therefore, SCK-CEN offers universities the possibility to perform (on a commercial basis) training exercises at its infrastructure consisting of two research reactors (BR1 and VENUS). Besides the organisation of training exercises in the framework of university courses, SCK-CEN also organizes theoretical courses in reactor physics for the education and training of nuclear reactor operators. It is indeed a very important subject to guarantee the safe operation of present and future nuclear reactors. In this framework, an understanding of the fundamental principles of nuclear reactor physics is also necessary for reactor operators. Therefore, the organisation of a basic Nuclear reactor physics course at the level of reactor operators in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The objectives this activity are: (1) to provide training and education activities in reactor physics for university students and (2) to organise courses in nuclear reactor physics for reactor operators

  11. System Definition Document: Reactor Data Necessary for Modeling Plutonium Disposition in Catawba Nuclear Station Units 1 and 2

    International Nuclear Information System (INIS)

    Ellis, R.J.

    2000-01-01

    The US Department of Energy (USDOE) has contracted with Duke Engineering and Services, Cogema, Inc., and Stone and Webster (DCS) to provide mixed-oxide (MOX) fuel fabrication and reactor irradiation services in support of USDOE's mission to dispose of surplus weapons-grade plutonium. The nuclear station units currently identified as mission reactors for this project are Catawba Units 1 and 2 and McGuire Units 1 and 2. This report is specific to Catawba Nuclear Station Units 1 and 2, but the details and materials for the McGuire reactors are very similar. The purpose of this document is to present a complete set of data about the reactor materials and components to be used in modeling the Catawba reactors to predict reactor physics parameters for the Catawba site. Except where noted, Duke Power Company or DCS documents are the sources of these data. These data are being used with the ORNL computer code models of the DCS Catawba (and McGuire) pressurized-water reactors

  12. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    1991-04-01

    The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

  13. Measurements and calculation of reactivity in the IEA-R1 nuclear reactor

    International Nuclear Information System (INIS)

    Ferreira, P.S.B.

    1988-01-01

    Techniques and experimentals procedures utilized in the measurement of some nuclear parameters related to reactivity are presented. Measurements of reactivity coefficients, such as void, temperature and power, and control rod worth were made in the IEA-R1 Research Reactor. The techniques used to perform the measurements were: i) stable period (control rod calibration), ii) inverse kinetics (digital reactivity meter), iii) aluminium slab insertion in the fuel element coolant channels (void reactivity), iv) nuclear reactor core temperature changes by means of the changes in the coolant systems of reactor core (isothermal reactivity coefficient) and v) by making perturbation in the core through the control rod motions (power reactivity coefficient and control rod calibration). By using the computer codes HAMMER, HAMMER-TECHNION and CITATION, the experiments realized in the IEA-R1 reactor were simulated. From this simulation, the theoretical reactivity parameters were estimated and compared with the respective experimental results. Furthermore, in the second fuel load of Angra-1 Nuclear Power Station, the IPEN-CNEN/SP digital reactivity - meter were used in the lower power test with the aim to assess the equipment performance. Among several tests, the reacticity-meter were used in parallel with a Westinghouse analogic reativimeter-meter) to measure the heat additiona point, critical boron concentration, control rod calibration, isothermal and moderator reactivity coefficient. These tests, and the results obtained by the digital reactivity-meter are described. The results were compared with those obtained by Westinghouse analogic reactivity meter, showing excellent agreement. (author) [pt

  14. Core calculations for the upgrading of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E.

    1998-01-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  15. Letter from Peter Davis

    International Nuclear Information System (INIS)

    Peter Davis

    2009-01-01

    This article begins with the writings of the efforts of certain groups who try to incorporate the information that is misunderstanding and conflict with modern scientific facts relating to the dangers of nuclear power plants and radioactive waste. At the same time these people try analyzes the idea that the solar panels and windmills can produce enough electricity to replace fossil fuels and nuclear power is not required to make. During the period of human life, the various forms of energy have been applied. In life and art that developed at this time, more energy is needed. Accordingly, the fourth-generation nuclear reactor with a production capacity of high temperature operation and is said to have special advantages for the supply of cheap energy and no pollution, is needed in developing countries.

  16. A Novel Event-Based Incipient Slip Detection Using Dynamic Active-Pixel Vision Sensor (DAVIS).

    Science.gov (United States)

    Rigi, Amin; Baghaei Naeini, Fariborz; Makris, Dimitrios; Zweiri, Yahya

    2018-01-24

    In this paper, a novel approach to detect incipient slip based on the contact area between a transparent silicone medium and different objects using a neuromorphic event-based vision sensor (DAVIS) is proposed. Event-based algorithms are developed to detect incipient slip, slip, stress distribution and object vibration. Thirty-seven experiments were performed on five objects with different sizes, shapes, materials and weights to compare precision and response time of the proposed approach. The proposed approach is validated by using a high speed constitutional camera (1000 FPS). The results indicate that the sensor can detect incipient slippage with an average of 44.1 ms latency in unstructured environment for various objects. It is worth mentioning that the experiments were conducted in an uncontrolled experimental environment, therefore adding high noise levels that affected results significantly. However, eleven of the experiments had a detection latency below 10 ms which shows the capability of this method. The results are very promising and show a high potential of the sensor being used for manipulation applications especially in dynamic environments.

  17. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  18. Strategies to prevent and reduce diabetes and obesity in Sacramento, California: the African American Leadership Coalition and University of California, Davis.

    Science.gov (United States)

    Ziegahn, Linda; Styne, Dennis; Askia, Joyce; Roberts, Tina; Lewis, Edward T; Edwards, Whitney

    2013-11-14

    Diabetes is one of the leading causes of illness and death for African Americans and people of African descent throughout the United States and in the city and county of Sacramento, California. The involvement of families and communities in developing prevention strategies can increase the likelihood that behavioral changes will be sustained. Three member organizations of the African American Leadership Coalition (AALC) entered into a partnership with the University of California, Davis (UC Davis) to engage families in developing a process to identify barriers to diabetes and obesity prevention and reduction, exchange strategies, and create action plans for prevention. The intervention comprised 3 phases: 1) coalition formation and training; 2) data collection, analysis, and dissemination of results; and 3) development of family and community action plans. Academic and community partners planned and implemented all project phases together. Sources of information about diabetes and obesity were primarily doctors and the Internet; barriers were related to lack of time needed to prepare healthy meals, high food costs, transportation to fresh markets, motivation around healthy habits, and unsafe environments. Action plans addressed behavioral change and family cohesion. The group discussion format encouraged mutual support and suggestions for better eating and physical exercise habits. This collaborative partnership model can strengthen existing group relationships or promote new affiliations that form the basis for future action coalitions. Participants worked both within and across groups to exchange information, stories of success and challenges, and specific health improvement strategies.

  19. Dose measurements in controlled area and laboratory of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Maretti Junior, Fausto; Alvarenga, Frederico Ladeia

    2005-01-01

    The workers doses in exposure areas to the radiation are so important for a Radioprotection Quality Program, as well as to guarantee the workers safety. For that it is necessary to raise the doses in the radiation areas, to obtain the accumulated dose in certain procedures for detailed studies. Several risings were accomplished to obtain the radiation levels in the areas where the workers are exposed due the operation of a research nuclear reactor and in the radioisotopes manipulation laboratories of a nuclear institute. The radiation levels and doses can be observed through graphs in the dependences of the Controlled Area 1 (AC-1) and the Reactor Laboratory. Those limits are in according of the CNEN-NE-3.01 work limits rules. The conclusion of the work allowed to demonstrate that the Laboratory of the Reactor and AC-1, have booth an effective radiological program with efficient operational practices that contributes with low doses to the workers. (author)

  20. Ageing Management Programme for the IEA-R1 Reactor in São Paulo, Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Ramanathan, L. V. [Institute of Energy and Nuclear Research (IPEN), National Nuclear Energy Commission (CNEN), São Paulo (Brazil)

    2014-08-15

    IEA-R1 is a swimming pool type reactor. It is moderated and cooled by light water and uses graphite and beryllium as reflector elements. First criticality was achieved on 16 September 1957, and the reactor is currently operating at 4.0 MW on a 64 h per week cycle. In 1996, a reactor ageing study was established to determine general deterioration of systems and components such as cooling towers, secondary cooling system, piping, pumps, specimen irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation, and safety system. The basic structure of the reactor from the original design has been maintained, but several improvements and modifications have been made over the years to various components, systems and structures. During the period 1996–2005 the reactor power was increased from 2 MW to 5 MW and the operational cycle from 8 h per day for 5 days a week to 120 h continuous per week, mainly to increase production of {sup 99}Mo. Prior to increasing reactor power, several modifications were made to the reactor system and its components. Simultaneously, a vigorous ageing management, inspection and modernization programme was put in place.

  1. Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

    Energy Technology Data Exchange (ETDEWEB)

    Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

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

    International Nuclear Information System (INIS)

    1985-03-01

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

  3. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    International Nuclear Information System (INIS)

    Faghihi, F.; Ramezani, E.; Yousefpour, F.; Mirvakili, S.M.

    2008-01-01

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation

  4. Level-1 probability safety assessment of the Iranian heavy water reactor using SAPHIRE software

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of); Nuclear Safety Research Center, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Ramezani, E. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of); Yousefpour, F. [Atomic Energy Organization of Iran (AEOI), Tehran (Iran, Islamic Republic of); Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, 71348-51153 Shiraz (Iran, Islamic Republic of)

    2008-10-15

    The main goal of this review paper is to analyze the total frequency of the core damage of the Iranian Heavy Water Research Reactor (IHWRR) compared with standard criteria and to determine the strengths and the weaknesses of the reactor safety systems towards improving its design and operation. The PSA has been considered for full-power state of the reactor and this article represents a level-1 PSA analysis using System Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) software. It is specifically designed to permit a listing of the potential accident sequences, compute their frequencies of occurrence and assign each sequence to a consequence. The method used for modeling the systems and accident sequences, is Large Fault Tree/Small Event Tree method. This PSA level-1 for IHWRR indicates that, based on conservative assumptions, the total frequency of accidents that would lead to core damage from internal initiating events is 4.44E-05 per year of reactor operation.

  5. Source term determination from subcritical multiplication measurements at Koral-1 reactor

    International Nuclear Information System (INIS)

    Blazquez, J.B.; Barrado, J.M.

    1978-01-01

    By using an AmBe neutron source two independent procedures have been settled for the zero-power experimental fast-reactor Coral-1 in order to measure the source term which appears in the point kinetical equations. In the first one, the source term is measured when the reactor is just critical with source by taking advantage of the wide range of the linear approach to critical for Coral-1. In the second one, the measurement is made in subcritical state by making use of the previous calibrated control rods. Several applications are also included such as the measurement of the detector dead time, the determinations of the reactivity of small samples and the shape of the neutron importance of the source. (author)

  6. Studies in fusion reactor technology. Final report, September 1, 1974--August 31, 1977

    International Nuclear Information System (INIS)

    Axtmann, R.C.; Perkins, H.K.

    1977-08-01

    Two independent measurements of hydrogen permeation through stainless steel at driving pressures in the range from 10 -6 to 1 Pa indicate that most extant predictions of tritium permeation through fusion reactors are probably overestimated grossly. A comprehensive analysis demonstrates that, given available structural materials, the prospects are negligible for the economic production of synthetic fuels via radiolytic reactions in fusion reactor systems

  7. Burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2011-01-01

    Full Text Available The burn-up dependent steady-state thermal hydraulic analysis of Pakistan research reactor-1, reference operating core, has been carried out utilizing standard computer codes WIMS/D4, CITATION, and RELAP5/MOD3.4. Reactor codes WIMS/D4 and CITATION have been used for the calculations of neutronic parameters including peaking factors and power profiles at different burn-up considering a xenon free core and also the equilibrium xenon values. RELAP5/MOD3.4 code was utilized for the determination of peak fuel centerline, clad and coolant temperatures to ensure the safety of the reactor throughout the cycle. The calculations reveal that the reactor is safe and no nucleate boiling will commence at any part of the core throughout the cycle and that the safety margin increases with burnup as peaking factors decrease.

  8. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    International Nuclear Information System (INIS)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

    2015-01-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  9. Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

    2015-07-01

    Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

  10. Measurement of thermal neutron flux spatial distribution in the IEA-R1 reactor core

    International Nuclear Information System (INIS)

    D'Utra Bitelli, U.

    1993-01-01

    This work presents the spatial thermal neutron flux in IEA-R1 reactor obtained by activation foils methods. These measurements were made in 27 fuel elements of the reactor core (165 B configuration). The results are important to compare with theoretical values, power calibration and safety analysis. (author)

  11. AKR-1 nuclear training reactor of Dresden Technical University turns twenty-five

    International Nuclear Information System (INIS)

    Hansen, W.

    2003-01-01

    Twenty-five years ago, in the night of July 27 to 28, 1978, the AKR-1 nuclear training reactor of the Dresden Technical University went critical for the first time and was commissioned. On the occasion of this anniversary, a colloquy was arranged with representatives from science, politics and industry, at which the reactor's history, the excellent achievements in research and training with the reactor, and the status and perspectives of this research facility were described. The AKR-1 had been built within the framework of the Nuclear Development Program of the then German Democratic Republic (GDR). The Nuclear Power Scientific Division of the Dresden Technical University had been entrusted with the responsibility, among other things, to train university personnel for the GDR Nuclear Power Program. The review by an expert group in 1996 of this plant had resulted in a recommendation in favor of long-term plant operation. A nuclear licensing procedure to this effect was initiated, and the necessary technical backfitting measures were implemented. The AKR-1 plant now equally serves for the specialized training of students and for research. (orig.) [de

  12. Measures aimed at enhancing safe operation of the Nigeria Research Reactor-1 (NIRR-1)

    International Nuclear Information System (INIS)

    Balogun, G.I.; Jonah, S.A.; Umar, I.M.

    2005-01-01

    Safety culture has been defined as 'that assembly of characteristics and attitudes in organizations and individuals which establishes that as an overriding priority, nuclear plant safety issues receive the attention warranted by their significance'. This paper briefly highlights efforts being made at the Centre for Energy Research and Training (CERT) towards realizing this broad objective as far as possible. To this end CERT realizes the need for instituted safety measures to reflect significant, site-specific peculiar characteristics of any generic reactor types. Consequently, standard procedures for pre-startup, startup and shutdown of NIRR-1 (a miniature neutron source reactor - MNSR) have been reviewed to reflect our local conditions and peculiarities. The review has revealed the need to incorporate important steps that impact on overall safety of the facility. For instance an interlocking system is being considered between NIRR-1 startup on the one hand and mandatory pre-startup measures on the other. Also a procedure has been put in place that would facilitate rapid response in the event of a rod-stuck-at-full-withdrawal incident. Furthermore, a program of automation of important analysis and design calculations of MNSRs is going on. Emphases are also placed, and deliberate efforts are being made, to ensure that a working atmosphere prevails that would foster the correct attitudinal approach to matters of reactor safety. A regime of constant dialogue and discussions amongst operating personnel has been factored into the overall operational program. (author)

  13. 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

  14. 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.

  15. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  16. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    International Nuclear Information System (INIS)

    Block, R.C.; Feiner, F.

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  17. Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 1, Sessions 1-5

    Energy Technology Data Exchange (ETDEWEB)

    Block, R.C.; Feiner, F. [comps.] [American Nuclear Society, La Grange Park, IL (United States)

    1995-09-01

    This document, Volume 1, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  18. Build-up of actinides in irradiated fuel rods of the ET-RR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Adib, M.; Naguib, K.; Morcos, H.N

    2001-09-01

    The content concentrations of actinides are calculated as a function of operating reactor regime and cooling time at different percentage of fuel burn-up. The build-up transmutation equations of actinides content in an irradiated fuel are solved numerically .A computer code BAC was written to operate on a PC computer to provide the required calculations. The fuel element of 10% {sup 235}U enrichment of ET-RR-1 reactor was taken as an example for calculations using the BAC code. The results are compared with other calculations for the ET-RR-1 fuel rod. An estimation of fissile build-up content of a proposed new fuel of 20% {sup 235}U enrichment for ET-RR-1 reactor is given. The sensitivity coefficients of build-up plutonium concentrations as a function of cross-section data uncertainties are also calculated.

  19. Calculation of the main neutron parameters of the IEA-R1 research reactor

    International Nuclear Information System (INIS)

    Ojima, Mario Katsuhiko

    1977-01-01

    The main neutron parameters of the research reactor IEA-R1 were calculated using computer programs to generate cross sections and criticality calculations. A calculation procedure based on the programs available in the Processing Center Data of IEA was established and centered in the HAMMER and CITATION system. A study was done in order to verify the validity and accuracy of the calculation method comparing the results with experimental data. Some operating parameters of the reactor, namely the distribution of neutron flux, the critical mass, the variation of the reactivity with the burning of fuel, and the dead time of the reactor were determined

  20. 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.

  1. 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

  2. Validation of the AZTRAN 1.1 code with problems Benchmark of LWR reactors; Validacion del codigo AZTRAN 1.1 con problemas Benchmark de reactores LWR

    Energy Technology Data Exchange (ETDEWEB)

    Vallejo Q, J. A.; Bastida O, G. E.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Xolocostli M, J. V.; Gomez T, A. M., E-mail: amhed.jvq@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The AZTRAN module is a computational program that is part of the AZTLAN platform (Mexican modeling platform for the analysis and design of nuclear reactors) and that solves the neutron transport equation in 3-dimensional using the discrete ordinates method S{sub N}, steady state and Cartesian geometry. As part of the activities of Working Group 4 (users group) of the AZTLAN project, this work validates the AZTRAN code using the 2002 Yamamoto Benchmark for LWR reactors. For comparison, the commercial code CASMO-4 and the free code Serpent-2 are used; in addition, the results are compared with the data obtained from an article of the PHYSOR 2002 conference. The Benchmark consists of a fuel pin, two UO{sub 2} cells and two other of MOX cells; there is a problem of each cell for each type of reactor PWR and BWR. Although the AZTRAN code is at an early stage of development, the results obtained are encouraging and close to those reported with other internationally accepted codes and methodologies. (Author)

  3. Current speed and direction, temperature and salinity collected from Moored Buoy in Davis Strait from 1987-09-05 to 1990-09-06 (NCEI Accession 0129882)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Volume, freshwater and heat transport through Davis Strait, the northern boundary of the Labrador Basin, were computed using a mooring array deployed for three...

  4. Fusion reactor technology studies. Final report for period August 1, 1972 - October 31, 1978

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Maynard, C.W.

    1984-04-01

    Major accomplishments for the period August 1, 1972 - October 31, 1978 include the publishing of four comprehensive fusion reactor conceptual design studies; experimental studies in the areas of radiation damage, plasma-wall interactions, superconducting magnets and 14-MeV neutron cross sections; development of the concepts of carbon curtains and ISSEC's for use in fusion reactors; development of a neutron and gamma heating computer code, a radioactivity and afterheat computer code and a neutral transport computer code; and studies in the areas of RF heating for tokamaks and resource assessment for fusion reactors

  5. Performance and Health Test Procedure for Grid Energy Storage Systems: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Baggu, Murali M [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Smith, Kandler A [National Renewable Energy Laboratory (NREL), Golden, CO (United States); Friedl, Andrew [San Diego Gas and Electric; Bialek, Thomas [San Diego Gas and Electric; Schimpe, Michael Robert [Technical University of Munich

    2017-07-27

    A test procedure to evaluate the performance and health of field installations of grid-connected battery energy storage systems (BESS) is described. Performance and health metrics captured in the procedures are: Round-trip efficiency, Standby losses, Response time/accuracy, and Useable Energy/ State of Charge at different discharge/charge rates over the system's lifetime. The procedures are divided into Reference Performance Tests, which require the system to be put in a test mode and are to be conducted in intervals, and Real-time Monitoring tests, which collect data during normal operation without interruption. The procedures can be applied on a wide array of BESS with little modifications and can thus support BESS operators in the management of BESS field installations with minimal interruption and expenditures.can be applied on a wide array of BESS with little modifications and can thus support BESS operators in the management of BESS field installations with minimal interruption and expenditures.

  6. PR-EDB: Power Reactor Embrittlement Data Base, version 1: Program description

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Taylor, B.J.

    1990-06-01

    Data concerning radiation embrittlement of pressure vessel steels in commercial power reactors have been collected form available surveillance reports. The purpose of this NRC-sponsored program is to provide the technical bases for voluntary consensus standards, regulatory guides, standard review plans, and codes. The data can also be used for the exploration and verification of embrittlement prediction models. The data files are given in dBASE 3 Plus format and can be accessed with any personal computer using the DOS operating system. Menu-driven software is provided for easy access to the data including curve fitting and plotting facilities. This software has drastically reduced the time and effort for data processing and evaluation compared to previous data bases. The current compilation of the Power Reactor Embrittlement Data base (PR-EDB, version 1) contains results from surveillance capsule reports of 78 reactors with 381 data points from 110 different irradiated base materials (plates and forgings) and 161 data points from 79 different welds. Results from heat-affected-zone materials are also listed. Electric Power Research Institute (EPRI), reactor vendors, and utilities are in the process of providing back-up quality assurance checks of the PR-EDB and will be supplementing the data base with additional data and documentation. 2 figs., 28 tabs

  7. Project management plan for the 105-C Reactor interim safe storage project. Revision 1

    International Nuclear Information System (INIS)

    Miller, R.L.

    1997-01-01

    In 1942, the Hanford Site was commissioned by the US Government to produce plutonium. Between 1942 and 1955, eight water-cooled, graphite-moderated reactors were constructed along the Columbia River at the Hanford Site to support the production of plutonium. The reactors were deactivated from 1964 to 1971 and declared surplus. The Surplus Production Reactor Decommissioning Project (BHI 1994b) will decommission these reactors and has selected the 105-C Reactor to be used as a demonstration project for interim safe storage at the present location and final disposition of the entire reactor core in the 200 West Area. This project will result in lower costs, accelerated schedules, reduced worker exposure, and provide direct benefit to the US Department of Energy for decommissioning projects complex wide. This project sets forth plans, organizational responsibilities, control systems, and procedures to manage the execution of the Project Management Plan for the 105-C Reactor Interim Safe Storage Project (Project Management Plan) activities to meet programmatic requirements within authorized funding and approved schedules. The Project Management Plan is organized following the guidelines provided by US Department of Energy Order 4700.1, Project Management System and the Richland Environmental Restoration Project Plan (DOE-RL 1992b)

  8. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, T L; Hellstrand, E; Londen, S O; Tiren, L I

    1965-08-15

    FR0 is a fast zero power reactor built for experiments in reactor physics. It is a split table machine containing vertical fuel elements. 120 kg of U{sup 235} are available as fuel, which is fabricated into metallic plates of 20 % enrichment. The control system comprises 5 spring-loaded safety elements and 3 + 1 elements for startup operations and power control. The reactor went critical in February 1964. The first assemblies studied were made up of undiluted fuel into a cylindrical and a spherical core, respectively, surrounded by a reflector made of copper. The present report describes some experiments made on these systems. Primarily, critical mass determinations, flux distribution measurements and studies of the conversion ratio are dealt with. The measured quantities have been compared with theoretical predictions using various transport theory programmes (DSN, TDC) and cross section sets. The experimental results show that the neutron spectrum in the copper reflector is softer than predicted, but apart from this discrepancy agreement with theory has generally been obtained.

  9. Neutron radiography in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Pugliesi, R.; Moraes, A.P.V. de; Yamazaki, I.M.; Freitas Acosta, C. de.

    1988-08-01

    Neutronradiography of several materials have been obtained at the IEA-R1 Nuclear Research Reactor (IPEN-CNEN/SP), by means of two conversion techniques: a) (n, α) at the beam-hole n 0 3 where a collimated thermal neutron beam, exposure area 4 cm x 8cm and flux at the sample 10 5 n/s cm 2 is obtained. The film used was the CN-85 cellulose nitrate coated with lithium tetraborate (conversor). The time irradiation of the film was 15 minutes and in following was eteched during 30 minutes in a NaOH(10%) aqueous solution at a constant temperature of 60 0 C.; b) (n,γ) by using an experimental arrangement installed in the botton of the pool of the reactor. The flux of the collimated neutron beam is 10 5 n/s/cm 2 at the sample and the conversion is made by means of a dysprozium sheet. The film used was Kodak T-5. The irradiation and the transfering time was 2 hours and 20 hours respectively. (author) [pt

  10. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients

  11. TRAC-PF1/MOD1: an advanced best-estimate computer program for pressurized water reactor thermal-hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Liles, D.R.; Mahaffy, J.H.

    1986-07-01

    The Los Alamos National Laboratory is developing the Transient Reactor Analysis Code (TRAC) to provide advanced best-estimate predictions of postulated accidents in light-water reactors. The TRAC-PF1/MOD1 program provides this capability for pressurized water reactors and for many thermal-hydraulic test facilities. The code features either a one- or a three-dimensional treatment of the pressure vessel and its associated internals, a two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field and solute tracking, flow-regime-dependent constitutive equation treatment, optional reflood tracking capability for bottom-flood and falling-film quench fronts, and consistent treatment of entire accident sequences including the generation of consistent initial conditions. The stability-enhancing two-step (SETS) numerical algorithm is used in the one-dimensional hydrodynamics and permits this portion of the fluid dynamics to violate the material Courant condition. This technique permits large time steps and, hence, reduced running time for slow transients.

  12. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia

    International Nuclear Information System (INIS)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E.

    2014-10-01

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  13. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  14. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

    2013-06-01

    (Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate

  15. 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.)

  16. Unitary theory of xenon instability in nuclear thermal reactors - 1. Reactor at 'zero power'

    Energy Technology Data Exchange (ETDEWEB)

    Novelli, A. (Politecnico di Milano (Italy). Centro Studi Nucleari E. Fermi)

    1982-01-01

    The question of nuclear thermal-reactor instability against xenon oscillations is widespread in the literature, but most theories, concerned with such an argument, contradict each other and, above all, they conflict with experimentally-observed instability at very low reactor power, i.e. without any power feedback. It is shown that, in any nuclear thermal reactor, xenon instability originates at very low power levels, and a very general stability condition is deduced by an extension of the rigorous, simple and powerful reduction of the Nyquist criterion, first performed by F. Storrer.

  17. Research reactors in Argentina

    International Nuclear Information System (INIS)

    Carlos Ruben Calabrese

    1999-01-01

    Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

  18. Measurements of reactivity of reactor G1

    International Nuclear Information System (INIS)

    Bernot, J.; Koechlin, J.C.; Portes, L.; Teste du Bailler, A.

    1957-01-01

    The various methods used during the physical study of the reactor G1 to determine the variations of the effective multiplication factor consecutive to a given change in the geometry of the multiplying medium, are presented and discussed. The comparison of the results obtained by these various methods has allowed their validity to be tested and precise conditions of use to be given. In the first part are presented the principles used and their ranges of validity. In the second part the experimental results are given, together with some indications on their comparison with theoretical estimations. (author) [fr

  19. Summary Report of Commercial reactor Criticality Data for Three Mile Island Unit 1

    International Nuclear Information System (INIS)

    Larry B. Wimmer

    2001-01-01

    The objective of the ''Summary Report of Commercial Reactor Criticality Data for Three Mile Island Unit I'' is to present the CRC data for the TMI-1 reactor. Results from the CRC evaluations will support the development and validation of the neutronics models used for criticality analyses involving commercial spent nuclear fuel. These models and their validation are discussed in the ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2000)

  20. Evaluation of off-type grasses in hybrid bermudagrass (Cynodon dactylon (L.) Pers. x C. transvaalensis Burtt-Davy) putting greens using genotyping-by-sequencing

    Science.gov (United States)

    Use of hybrid ultradwarf bermudagrasses (UDBG; Cynodon dactylon (L.) Pers. x C. transvaalensis Burtt-Davy) on golf course putting greens is increasing in the southern United States. However, off-type grasses within many putting surfaces have been observed. To explore the genetic variation among UD...

  1. Modifications in the operational conditions of the IEA-R1 reactor under continuous 48 hours operation

    International Nuclear Information System (INIS)

    Moreira, Joao Manoel Losada; Frajndlich, Roberto

    1995-01-01

    This work shows the required changes in the IEA-R1 reactor for operation at 2 Mw, 48 hours continuously. The principal technical change regards the operating conditions of the reactor, namely, the required excess reactivity which now will amount to 4800 pcm in order to compensate the Xe poisoning at equilibrium at 2 Mw. (author). 6 refs, 1 fig, 1 tab

  2. Advance High Temperature Inspection Capabilities for Small Modular Reactors: Part 1 - Ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Bond, Leonard J. [Iowa State Univ., Ames, IA (United States); Bowler, John R. [Iowa State Univ., Ames, IA (United States)

    2017-08-30

    The project objective was to investigate the development non-destructive evaluation techniques for advanced small modular reactors (aSMR), where the research sought to provide key enabling inspection technologies needed to support the design and maintenance of reactor component performance. The project tasks for the development of inspection techniques to be applied to small modular reactor are being addressed through two related activities. The first is focused on high temperature ultrasonic transducers development (this report Part 1) and the second is focused on an advanced eddy current inspection capability (Part 2). For both inspection techniques the primary aim is to develop in-service inspection techniques that can be carried out under standby condition in a fast reactor at a temperature of approximately 250°C in the presence of liquid sodium. The piezoelectric material and the bonding between layers have been recognized as key factors fundamental for development of robust ultrasonic transducers. Dielectric constant characterization of bismuth scantanate-lead titanate ((1-x)BiScO3-xPbTiO3) (BS-PT) has shown a high Curie temperature in excess of 450°C , suitable for hot stand-by inspection in liquid metal reactors. High temperature pulse-echo contact measurements have been performed with BS-PT bonded to 12.5 mm thick 1018-low carbon steel plate from 20C up to 260 C. High temperature air-backed immersion transducers have been developed with BS-PT, high temperature epoxy and quarter wavlength nickel plate, needed for wetting ability in liquid sodium. Ultrasonic immersion measurements have been performed in water up to 92C and in silicone oil up to 140C. Physics based models have been validated with room temperature experimental data with benchmark artifical defects.

  3. Analysis of core melt accident in Fukushima Daiichi-Unit 1 nuclear reactor

    International Nuclear Information System (INIS)

    Tanabe, Fumiya

    2011-01-01

    In order to obtain a profound understanding of the serious situation in Unit 1 and Unit 2/3 reactors of Fukushima Daiichi Nuclear Power Station (hereafter abbreviated as 1F1 and 1F2/3, respectively), which was directly caused by tsunami due to a huge earthquake on 11 March 2011, analyses of severe core damage are performed. In the present report, the analysis method and 1F1 analysis are described. The analysis is essentially based on the total energy balance in the core. In the analysis, the total energy vs. temperature curve is developed for each reactor, which is based on the estimated core materials inventory and material property data. Temperature and melt fraction are estimated by comparing the total energy curve with the total stored energy in the core material. The heat source is the decay heat of fission products and actinides together with reaction heat from the zirconium steam reaction. (author)

  4. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...

  5. Pressure Sores and Systemic Inflammatory Response Syndrome: UC Davis Quality Improvement Initiative.

    Science.gov (United States)

    Jairam, Abhishek; Song, Ping; Patel, Nirav B; Wong, Michael S

    2018-05-01

    The National Pressure Ulcer Advisory Panel estimates pressure sore care to approach $11 billion annually. It is not uncommon for these patients to present to the emergency department (ED) with a chief concern of a pressure sore, while concurrently carrying an undiagnosed infectious process that is the culprit for the acute presentation, rather than the chronic pressure injury. We aim to identify patients who met systemic inflammatory response syndrome (SIRS) criteria at ED presentation who were referred to plastic and reconstructive surgery for pressure sore debridement prior to a complete medical workup. We hypothesize that a restructuring of the ED triaging system would help conserve hospital resources, reduce costs of pressure sore management, and improve patient care and outcomes by first treating primary, underlying pathologies. This is a retrospective chart review of 36 patients who presented to the University of California, Davis Medical Center Emergency Department with a pressure sore and met SIRS criteria, but obtained a plastic surgery consult prior to a full medical workup. We defined SIRS based on standardized criteria: temperature greater than 100.4°F or less than 96.8°F, pulse rate greater than 90 beats/min, respiratory rate greater than 20 breaths/min or PaCO2 less than 32 mm Hg, white blood cell count greater than 12,000, less than 4000, or greater than 10% bands. Fifty percent of patients (18/36) met SIRS criteria at ED presentation for their pressure sores. Of these SIRS patients, 9 (50%) had a diagnosis of urinary tract infection or urosepsis, 6 (33.3%) had sepsis of undefined origin, and 3 (16.7%) had other diagnoses such as osteomyelitis or acute respiratory distress syndrome. Half of patients consulted while in the University of California, Davis Medical Center Emergency Department with pressure sores met SIRS criteria and received a plastic and reconstructive surgery consult prior to a full medical workup. We propose a new algorithm for

  6. Reactor of the XXI century

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.; Solov'ev, Yu.A.

    2001-01-01

    The advantages of both molten salt reactors (MSR) and homogenous molten salt reactors (HMSR) are illuminated. It is noted that the MSR possess accident probability A=10 -6 1/reactor.years and the HMSR with integral configuration has A=10 -7 1/reactor.years. The methods for these reactors technological problems solution - tritium removal, salt melt circulation and capacity pick up - are discussed

  7. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  8. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  9. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  10. Extension of cycle 8 of Angra-1 reactor, optimization of electric power generation reduction

    International Nuclear Information System (INIS)

    Miranda, Anselmo Ferreira; Moreira, Francisco Jose; Valladares, Gastao Lommez

    2000-01-01

    The main objective of extending fuel cycle length of Angra-1 reactor, is in fact of that each normal refueling are changed about 40 fuel elements of the reactor core. Considering that these elements do not return for the reactor core, this procedure has became possible a more gain of energy of these elements. The extension consists in, after power generation corresponding to a cycle burnup of 13700 MWD/TMU or 363.3 days, to use the reactivity gain by reduction of power and temperature of primary system for power generation in a low energy patamar

  11. Power Trip Set-points of Reactor Protection System for New Research Reactor

    International Nuclear Information System (INIS)

    Lee, Byeonghee; Yang, Soohyung

    2013-01-01

    This paper deals with the trip set-point related to the reactor power considering the reactivity induced accident (RIA) of new research reactor. The possible scenarios of reactivity induced accidents were simulated and the effects of trip set-point on the critical heat flux ratio (CHFR) were calculated. The proper trip set-points which meet the acceptance criterion and guarantee sufficient margins from normal operation were then determined. The three different trip set-points related to the reactor power are determined based on the RIA of new research reactor during FP condition, over 0.1%FP and under 0.1%FP. Under various reactivity insertion rates, the CHFR are calculated and checked whether they meet the acceptance criterion. For RIA at FP condition, the acceptance criterion can be satisfied even if high power set-point is only used for reactor trip. Since the design of the reactor is still progressing and need a safety margin for possible design changes, 18 MW is recommended as a high power set-point. For RIA at 0.1%FP, high power setpoint of 18 MW and high log rate of 10%pp/s works well and acceptance criterion is satisfied. For under 0.1% FP operations, the application of high log rate is necessary for satisfying the acceptance criterion. Considering possible decrease of CHFR margin due to design changes, the high log rate is suggested to be 8%pp/s. Suggested trip set-points have been identified based on preliminary design data for new research reactor; therefore, these trip set-points will be re-established by considering design progress of the reactor. The reactor protection system (RPS) of new research reactor is designed for safe shutdown of the reactor and preventing the release of radioactive material to environment. The trip set point of RPS is essential for reactor safety, therefore should be determined to mitigate the consequences from accidents. At the same time, the trip set-point should secure margins from normal operational condition to avoid

  12. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  13. The Osiris reactor. Descriptive report - Volume 1 - text

    International Nuclear Information System (INIS)

    1969-05-01

    Osiris is a pool type reactor with a 70 MW thermal power. Its main purpose is to irradiate under high flows of neutrons the materials of which future nuclear power stations are made. This report proposes a description of this pool reactor. A first part describes the functional aspects and general characteristics of all installations which are in principle definitely defined (premises, irradiation and experimentation equipment, water circuits, power supply, venting, controls). The second part addresses elements which are likely to be changed, and more particularly the reactor core: fuel elements and controls (uranium and boron load in different fuel element generations, experimental locations within the core), neutron transport aspects (calculation and experiment), and thermal aspects (power generation and removal) of the pile). The third part addresses the operation: operation cycles, stops, exploitation organisation [fr

  14. CANDU reactors with reactor grade plutonium/thorium carbide fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sahin, Suemer [Atilim Univ., Ankara (Turkey). Faculty of Engineering; Khan, Mohammed Javed; Ahmed, Rizwan [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan); Gazi Univ., Ankara (Turkey). Faculty of Technology

    2011-08-15

    Reactor grade (RG) plutonium, accumulated as nuclear waste of commercial reactors can be re-utilized in CANDU reactors. TRISO type fuel can withstand very high fuel burn ups. On the other hand, carbide fuel would have higher neutronic and thermal performance than oxide fuel. In the present work, RG-PuC/ThC TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 60%. The fuel compacts conform to the dimensions of sintered CANDU fuel compacts are inserted in 37 zircolay rods to build the fuel zone of a bundle. Investigations have been conducted on a conventional CANDU reactor based on GENTILLYII design with 380 fuel bundles in the core. Three mixed fuel composition have been selected for numerical calculation; (1) 10% RG-PuC + 90% ThC; (2) 30% RG-PuC + 70% ThC; (3) 50% RG-PuC + 50% ThC. Initial reactor criticality values for the modes (1), (2) and (3) are calculated as k{sub {infinity}}{sub ,0} = 1.4848, 1.5756 and 1.627, respectively. Corresponding operation lifetimes are {proportional_to} 2.7, 8.4, and 15 years and with burn ups of {proportional_to} 72 000, 222 000 and 366 000 MW.d/tonne, respectively. Higher initial plutonium charge leads to higher burn ups and longer operation periods. In the course of reactor operation, most of the plutonium will be incinerated. At the end of life, remnants of plutonium isotopes would survive; and few amounts of uranium, americium and curium isotopes would be produced. (orig.)

  15. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  16. Monitoring of primary circuit and reactor of NPP A-1

    International Nuclear Information System (INIS)

    Prazska, M.; Majersky, M.; Rezbarik, J.; Sekely, S.; Vozarik, P.; Walthery, R.; Stuller, P.

    2005-01-01

    Nuclear Power Plant A-1 in Jaslovske Bohunice was commissioned in 1972. Heavy water moderated, carbon dioxide cooled channel type reactor was shut down after two accidents in 1977. During more serious second accident, the reduced coolant flow caused local overheating of the fuel and consequent damage/melting of the fuel channel. Both accidents had led to the damage of several fuel assemblies with extensive local damage of fuel claddings. As a consequence, the main cooling circuit was significantly contaminated by fission products and long-life alpha nuclides. The detailed monitoring of dose rates, smearable contamination and sampling of contamination was performed. Extended monitoring in reacto vessel, primary circuit pipes, turbo-compressors, steam generators, main valves, gas tanks and also heavy water system with collectors, coolers, distilling and purification station, pumps and valves was done. Appropriate devices and procedures for the monitoring and examination of the installations were prepared and applied. Obtained results will serve for the future planning of the decontamination and decommissioning works. The 3-D model of the reactor that had been developed as part of this Project proved invaluable for orientation, visualisation, planning and analysis of results. Dose rates were measured in the technological channels from the reactor hall floor to the bottom of the hot gas chamber in decrements of 1 m and 0.5 m. The highest absolute values of dose rates were found in channels located in the middle of the reactor (up to 1900 mGy/h in the active zone region). It is estimated that the total contaminated area of primary circuit equipment (pipework, steam generators and turbo-compressors) is some 48 000 m 2 . It follows that the total gamma contamination is of the order of 10 14 to 10 15 Bq and total alpha contamination 10 11 to 10 13 Bq. The total amount of deposits in the gas circuit is about 14.3 tons. (authors)

  17. Characterisation of reactor control rod drives. Specification 1-6. Reaktorstellstabantriebe. Typenblaetter 1-6

    Energy Technology Data Exchange (ETDEWEB)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN).

  18. The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

    International Nuclear Information System (INIS)

    Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

    2004-01-01

    The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

  19. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  20. Development and characterization of genomic SSR markers in Cynodon transvaalensis Burtt-Davy.

    Science.gov (United States)

    Tan, Chengcheng; Wu, Yanqi; Taliaferro, Charles M; Bell, Greg E; Martin, Dennis L; Smith, Mike W

    2014-08-01

    Simple sequence repeat (SSR) markers are a major molecular tool for genetic and genomic research that have been extensively developed and used in major crops. However, few are available in African bermudagrass (Cynodon transvaalensis Burtt-Davy), an economically important warm-season turfgrass species. African bermudagrass is mainly used for hybridizations with common bermudagrass [C. dactylon var. dactylon (L.) Pers.] in the development of superior interspecific hybrid turfgrass cultivars. Accordingly, the major objective of this study was to develop and characterize a large set of SSR markers. Genomic DNA of C. transvaalensis '4200TN 24-2' from an Oklahoma State University (OSU) turf nursery was extracted for construction of four SSR genomic libraries enriched with [CA](n), [GA](n), [AAG](n), and [AAT](n) as core repeat motifs. A total of 3,064 clones were sequenced at the OSU core facility. The sequences were categorized into singletons and contiguous sequences to exclude redundancy. From the two sequence categories, 1,795 SSR loci were identified. After excluding duplicate SSRs by comparison with previously developed SSR markers using a nucleotide basic local alignment tool, 1,426 unique primer pairs (PPs) were designed. Out of the 1,426 designed PPs, 981 (68.8 %) amplified alleles of the expected size in the donor DNA. Polymorphisms of the SSR PPs tested in eight C. transvaalensis plants were 93 % polymorphic with 544 markers effective in all genotypes. Inheritance of the SSRs was examined in six F(1) progeny of African parents 'T577' × 'Uganda', indicating 917 markers amplified heritable alleles. The SSR markers developed in the study are the first large set of co-dominant markers in African bermudagrass and should be highly valuable for molecular and traditional breeding research.

  1. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  2. Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

    Energy Technology Data Exchange (ETDEWEB)

    Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

    2016-06-15

    Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

  3. Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

    International Nuclear Information System (INIS)

    Duque O, J.; Velez A, L.H.

    1987-01-01

    Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it

  4. Fuzzy Logic-Based Operation of Battery Energy Storage Systems (BESSs for Enhancing the Resiliency of Hybrid Microgrids

    Directory of Open Access Journals (Sweden)

    Akhtar Hussain

    2017-02-01

    Full Text Available The resiliency of power systems can be enhanced during emergency situations by using microgrids, due to their capability to supply local loads. However, precise prediction of disturbance events is very difficult rather the occurrence probability can be expressed as, high, medium, or low, etc. Therefore, a fuzzy logic-based battery energy storage system (BESS operation controller is proposed in this study. In addition to BESS state-of-charge and market price signals, event occurrence probability is taken as crisp input for the BESS operation controller. After assessing the membership levels of all the three inputs, BESS operation controller decides the operation mode (subservient or resilient of BESS units. In subservient mode, BESS is fully controlled by an energy management system (EMS while in the case of resilient mode, the EMS follows the commands of the BESS operation controller for scheduling BESS units. Therefore, the proposed hybrid microgrid model can operate in normal, resilient, and emergency modes with the respective objective functions and scheduling horizons. Due to the consideration of resilient mode, load curtailment can be reduced during emergency operation periods. Numerical simulations have demonstrated the effectiveness of the proposed strategy for enhancing the resiliency of hybrid microgrids.

  5. Calculations of Changes in Reactivity during some regular periods of operation of JEN-1 MOD Reactor

    International Nuclear Information System (INIS)

    Alcala Ruiz, F.

    1973-01-01

    By a Point-Reactor model and Perturbation Theory, changes in reactivity during some regular operating periods of JEN-1 MOD Reactor have been calculated and compared with available measured values. they were in good agreement. Also changes in reactivity have been calculated during operations at higher power levels than the present one, concluding some practical consequences for the case of increasing the present power of this reactor. (Author)

  6. Characterisation of reactor control rod drives. Specification 1-6

    International Nuclear Information System (INIS)

    1975-03-01

    The committee 'Kernreaktorregelung' of VDI/VDE-Gesellschaft Mess- und Regelungstechnik has developed 6 specifications (Typenblaetter) of reactor control rod drives. The specifications are aimed at giving engineers in reactor control systems an outline concerning the function as well as some construction characteristics. (orig./LN) [de

  7. Developing maintainability in controlled thermonuclear reactors. Progress report, October 1, 1977--April 30, 1978

    International Nuclear Information System (INIS)

    Zahn, H.S.

    1977-05-01

    During the period 1 October 1977 through 30 April 1978 the study has completed work on Task 6, Candidate Reference Systems. Four candidate reference systems have been defined. These are based on the conceptual designs of the UWMAK-III, the General Atomic Company Demonstration Power Reactor, the Oak Ridge National Laboratory Cassette defined in the Demonstration Power Study and the Culham laboratory Mark II Reactors. These reactor concepts are normalized to 3000 MW/sub th/ and near minimum cost of electricity. In addition, designs of four major subsystems have been selected and defined for application to these reactors. These include a primary coolant system, primary and secondary vacuum zone systems, the neutral beam injection system and the magnetic field system. These magnet systems are unique to each reactor. The cases for which maintenance plans are being developed in Task 7 have been selected to allow evaluation of design features, particularly the vacuum wall locations, and the impacts of unscheduled and contact maintenance of subsystems on the cost of electricity

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  9. TU Electric reactor physics model verification: Power reactor benchmark

    International Nuclear Information System (INIS)

    Willingham, C.E.; Killgore, M.R.

    1988-01-01

    Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

  10. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  11. Supervisory system to monitor the neutron flux of the IPR-R1 TRIGA research reactor at CDTN

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Tello, Cledola Cassia Oliveira

    2009-01-01

    The IPR-R1 TRIGA Mark I nuclear research reactor at the Nuclear Technology Development Center - CDTN (Belo Horizonte) is a pool type reactor. It was designed for research, training and radioisotope production. The International Atomic Energy Agency- IAEA - recommends the use of friendly interfaces for monitoring and controlling the operational parameters of nuclear reactors. This paper reports the activities for implementing a supervisory system, using LabVIEW software, with the purpose to provide the IPR-R1 TRIGA research reactor with a modern, safe and reliable system to monitor the time evolution of the power of its core. The use of the LabVIEW will introduce modern techniques, based on electronic processor and visual interface in video monitor, substituting the mechanical strip chart recorders (ink-pen drive and paper) that monitor the current neutrons flux, which is proportional to the thermal power supplied by reactor core. The main objective of the system will be to follow the evolution of the neutronic flux originated in the Linear and Logarithmic channels. A great advantage of the supervisory software nowadays, in relation to computer programs currently used in the facility, is the existence of new resources such as the data transmission and graphical interfaces by net, grid lines display in the graphs, and resources for real time reactor core video recordings. The considered system could also in the future be optimized, not only for data acquisition, but also for the total control of IPR-R1 TRIGA reactor(author)

  12. Probabilistic risk analysis of Angra-1 reactor

    International Nuclear Information System (INIS)

    Spivak, R.C.; Collussi, I.; Silva, M.C. da; Onusic Junior, J.

    1986-01-01

    The first phase of probabilistic study for safety analysis and operational analysis of Angra-1 reactor is presented. The study objectives and uses are: to support decisions about safety problems; to identify operational and/or project failures; to amplify operator qualification tests to include accidents in addition to project base; to provide informations to be used in development and/or review of operation procedures in emergency, test and maintenance procedures; to obtain experience for data collection about abnormal accurences; utilization of study results for training operators; and training of evaluation and reliability techniques for the personnel of CNEN and FURNAS. (M.C.K.) [pt

  13. Fusion reactor design and technology 1986. V. 1

    International Nuclear Information System (INIS)

    1987-01-01

    The first volume of the Proceedings of the Fourth Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology organized by the IAEA (Yalta, 26 May - 6 June 1986) includes 36 papers devoted to the following topics: fusion programmes (3 papers), tokamaks (15 papers), non-tokamak reactors and open systems (9 papers), inertial confinement concepts (5 papers), fission-fusion hybrids (4 papers). Each of these papers has a separate abstract. Refs, figs and tabs

  14. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    Ashoub, N.; Saleh, H.

    1995-01-01

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  15. Integral tightness measurements at the Paks-1 nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taubner, R.; Techy, Z. (Villamosenergiaipari Kutato Intezet, Budapest (Hungary))

    1983-01-01

    The containment system experiments of the Paks-1 nuclear reactor are described. The integrated tightness measurements of the hermetic system were completed in 1982. The principles and methods and the evaluation of the results of the measurements are discussed. Some features of the filtration characteristics are demonstrated using relative values and a method enabling the description of the physical contents of the characteristics by flow technical functions is outlined.

  16. Molten-salt reactor strategies viewed from fuel conservation effect, (1)

    International Nuclear Information System (INIS)

    Furuhashi, Akira

    1976-01-01

    Saving of material requirements in the long-term fuel cycle is studied by introducing molten-salt reactors with good neutron economy into a projection of nuclear generating capacity in Japan. In this first report an examination is made on the effects brought by the introduction of molten-salt converter reactors starting with Pu which are followed by 233 U breeders of the same type. It is shown that the sharing of some Pu in the light water- and fast breeder-reactor system with molten-salt reactors provides a more rapid transition to the self-supporting, breeding cycle than the simple fast breeding system, thus leading to an appreciable fuel conservation. Considerations are presented on the strategic repartition of generating capacity among reactor types and it is shown that all of the converted 233 U should be promptly invested to molten-salt breeders to quickly establish the dual breeding system, instead of recycling to converters themselves. (auth.)

  17. Security devices and experiment facilities at ENEA TRIGA RC-1 reactor

    International Nuclear Information System (INIS)

    Bianchi, P.; Festinesi, A.; Santoro, E.; Tardani, G.; Magli, M.; Reis, G.

    1990-01-01

    RC-1 TRIGA operating exercise staff has produced some auxiliary security devices. These are the neutron source automatic handling device, irradiated samples rabbit connection rotating rack, and auxiliary equipment for transferring hot fuel elements. The reactor electronic control instrumentation system includes various instrumentation channels, the operating capability of which must be verified by the licensee as per Italian regulations. In order to obtain automatic and repeatable operations, TEMAV designed and constructed a remotely-driven source transfer device, based on requirements, performance specifications and technical data supplied by ENEA-TIB. The pneumatic irradiating system for short lived materials allows extraction of radiated samples in a time no longer than 4 seconds. To optimize the system, both as to operability and health protection, a specific rotating rack for the connection of irradiated samples with pneumatic transfer (RABBIT) was produced. To permit 1 MW hot fuel element storage in pits it is necessary to remove hot 100 KW fuel elements and transfer them to a re-treatment plant. Feasibility studies showed the impossibility of using heavy trucks inside the reactor hall. To avoid problems trucks are left outside the reactor hall and only the PEGASO container is removed with a special device that runs on rails. Movement from Rail truck is assured by an electromotor driving pull device and security cable

  18. Qualification process of dispersion fuels in the IEAR1 research reactor

    International Nuclear Information System (INIS)

    Domingos, D.B.; Silva, A.T.; Silva, J.E.R.

    2010-01-01

    Neutronic, thermal-hydraulics and accident analysis calculations were developed to estimate the safety of a miniplate irradiation device (MID) to be placed in the IEA-R1 reactor core. The irradiation device will be used to receive miniplates of U 3 O 8 -Al and U 3 Si 2 -Al dispersion fuels, LEU type (19,9% of 235 U) with uranium densities of, respectively, 3.0 gU/cm 3 and 4.8 gU/cm 3 . The fuel miniplates will be irradiated to nominal 235 U burnup levels of 50% and 80%, in order to qualify the above high-density dispersion fuels to be used in the Brazilian Multipurpose Reactor (RMB), now in the conception phase. For the neutronic calculation, the computer codes CITATION and TWODB were utilized. The computer code FLOW was used to calculate the coolant flow rate in the irradiation device, allowing the determination of the fuel miniplate temperatures with the computer model MTRCR-IEA-R1. A postulated Loss of Coolant Accident (LOCA) was analyzed with the computer code LOSS and TEMPLOCA, allowing the calculation of the fuel miniplate temperatures after the reactor pool draining. This paper also presents a system designed for fuel swelling evaluation. The determination of the fuel swelling will be performed by means of the fuel miniplate thickness measurements along the irradiation time. (author)

  19. HAV-1-A multipurpose multimonitor for reactor neutron flux characterization

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Alvarez, I.; Herrera, E.; Lima, L.; Tores, J.; Lopez, M.C.; Ixquiac, M.

    1996-01-01

    A simple method non-solid multi monitor HAV-1 for the systematic evaluation of reactor neutron flux parameters for K o neutron activation analysis is presented. Solution of Au, Zr, Co, Zn, Sn, U and Th (deposited in filter paper) are used to study the parameters alpha and f. Dissolved Lu is used to neutron temperature (Tn) determination, according to the Wescott's formalism. A multipurpose multi monitor HAV-1 preparation, certification and evaluations presented

  20. Membrane-aerated biofilm reactor for the removal of 1,2-dichloroethane by Pseudomonas sp strain DCA1

    NARCIS (Netherlands)

    Hage, J.C.; Houten, R.T.; Tramper, J.; Hartmans, S.

    2004-01-01

    A membrane-aerated biofilm reactor (MBR) with a biofilm of Pseudomonas sp. strain DCA1 was studied for the removal of 1,2-dichloroethane (DCA) from water. A hydrophobic membrane was used to create a barrier between the liquid and the gas phase. Inoculation of the MBR with cells of strain DCA1 grown