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Sample records for pacific gas diablo canyon-1 reactor

  1. 2010 Pacific Gas and Electric Diablo Canyon Power Plant (DCPP): Diablo Canyon, CA Central Coast

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Diablo Canyon Power Plant (DCPP) LiDAR and Imagery datasets are comprised of three separate LiDAR surveys: Diablo Canyon (2010), Diablo Canyon (2010), and San...

  2. 75 FR 75704 - Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 And 2); Notice of...

    Science.gov (United States)

    2010-12-06

    ... NUCLEAR REGULATORY COMMISSION [Docket Nos. 50-275-LR; 50-323-LR] Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 And 2); Notice of Appointment of Adjudicatory Employee... Seismologist, Office of Nuclear Material Safety and Safeguards, has been appointed as a Commission adjudicatory...

  3. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1984-07-01

    Supplement 27 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for a license to operate Diablo Canyon Nuclear Power Plant, Unit 1 (Docket No. 50-275), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses the revisions to the license conditions and to the Technical Specifications as they relate to Amendment 10 to Diablo Canyon, Unit 1 Facility Operating License, DPR-76

  4. Safety-evaluation report related to the operation of Diablo Canyon Nuclear Power Plants, Units 1 and 2. Docket Nos. 50-275 and 50-323. Supplement No. 18

    International Nuclear Information System (INIS)

    1983-08-01

    Supplement 18 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the verification effort for Diablo Canyon Unit 1 that was performed between November 1981 and the present in response to Commission Order CLI-81-30 and an NRC letter to the licensee

  5. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1991-06-01

    Supplement 34 to the Safety Evaluation Report for the application by Pacific Gas and Electric Company (PG ampersand E) for licenses to operate Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2 (Docket Nos. 50-275 and 50-323, respectively) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement documents the NRC staff review of the Long-Term Seismic Program conducted by PG ampersand E in response to License Condition 2.C.(7) of Facility Operating License DPR-80, the Diablo Canyon Unit 1 operating license. 111 refs., 20 figs., 31 tabs

  6. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1983-12-01

    Supplement 20 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Unit 1 and Unit 2 (Docket Nos. 50-275 and 50-323), has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports on the verification effort for Diablo Canyon Unit 1 that was performed between November 1981 and the present in response to Commission Order CLI-81-30 and an NRC letter of November 19, 1981 to the licensee. Specifically, Supplement 20 addresses those issues and other matters identified in Supplements 18 and 19 that must be resolved prior to Unit 1 achieving criticality and operating at power levels up to 5% of rated full power. This SER Supplement applies only to Diablo Canyon Unit 1

  7. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 26

    International Nuclear Information System (INIS)

    1984-07-01

    Supplement 26 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. The supplement reports on the status of the staff's investigation, inspection and evaluation of those allegations or concerns that have been identified to the NRC as of July 8, 1984. The report specifically addresses those allegations which the staff determined must be satisfactorily resolved prior to full power operation of Diablo Canyon Unit 1

  8. 77 FR 7211 - Pacific Gas and Electric Company, Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2012-02-10

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 72-26; NRC-2011-0110] Pacific Gas and Electric Company...) issued NRC Materials License No. SNM-2511 to the Pacific Gas and Electric Company (PG&E) for the Diablo.... 5. TS 3.1.2, ``Spent Fuel Storage Cask (SFSC) Heat Removal System,''--revise to allow the HI-STORM...

  9. 2013 Pacific Gas and Electric Diablo Canyon Power Plant (DCPP): San Simeon, CA Central Coast

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Diablo Canyon Power Plant (DCPP) LiDAR and Imagery datasets are comprised of three separate LiDAR surveys: Diablo Canyon (2010), Los Osos (2011), and San Simeon...

  10. 2011 Pacific Gas and Electric Diablo Canyon Power Plant (DCPP): Los Osos, CA Central Coast

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Diablo Canyon Power Plant (DCPP) LiDAR and Imagery datasets are comprised of three separate LiDAR surveys: Diablo Canyon (2010), Los Osos (2011), and San Simeon...

  11. Diablo Canyon

    International Nuclear Information System (INIS)

    Bindon, F.J.L.

    1986-01-01

    The paper traces the history of Diablo Canyon nuclear power station, California, which took 18 years to reach full-power testing from the planning stage. The major delays during the construction are outlined, as well as the costs of Diablo Canyon. (UK)

  12. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1984-02-01

    Supplement 17 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report and the previous supplements

  13. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 33

    International Nuclear Information System (INIS)

    1986-05-01

    Supplement 33 to the Safety Evaluation Report for the Pacific Gas and Electric Company's Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. The supplement reports on the status of the staff's investigation, inspection, and evaluation of allegations and concerns that have been identified to the NRC through March 1986. The report includes a complete listing of all allegations and concerns, indicating the status of their resolution. The NRC staff concludes that the technical issues raised in the allegations with regard to the design, construction, and safe operation of Diablo Canyon Units 1 and 2 have been satisfactorily resolved and no further action is required

  14. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1984-06-01

    Supplement 23 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. This supplement addresses the applicant's requests for approval of 22 deviations from the requirements of Section III.G of Appendix R of Title 10 of the Code of Federal Regulations Part 50

  15. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 28

    International Nuclear Information System (INIS)

    1985-04-01

    Supplement 28 to the Safety Evaluation Report for the Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323) has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. The supplement reports on the status of the staff's investigation, inspection, and evaluation of those allegations or concerns that have been identified to the NRC as of March 1, 1985

  16. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 25

    International Nuclear Information System (INIS)

    1984-07-01

    Supplement 25 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Unit 1 and Unit 2 (Docket Nos. 50-275 and 50-323) has been prepared by the Office of Nuclear Reactor Regulation (NRR) of the US Nuclear Regulatory Commission. This supplement reports on the staff's inspection and evaluation efforts on the matter of piping and piping supports as related to the seven technical license conditions in an Order Modifying License issued by NRR on April 18, 1984

  17. Review of the Diablo Canyon probabilistic risk assessment

    International Nuclear Information System (INIS)

    Bozoki, G.E.; Fitzpatrick, R.G.; Bohn, M.P.; Sabek, M.G.; Ravindra, M.K.; Johnson, J.J.

    1994-08-01

    This report details the review of the Diablo Canyon Probabilistic Risk Assessment (DCPRA). The study was performed under contract from the Probabilistic Risk Analysis Branch, Office of Nuclear Reactor Research, USNRC by Brookhaven National Laboratory. The DCPRA is a full scope Level I effort and although the review touched on all aspects of the PRA, the internal events and seismic events received the vast majority of the review effort. The report includes a number of independent systems analyses sensitivity studies, importance analyses as well as conclusions on the adequacy of the DCPRA for use in the Diablo Canyon Long Term Seismic Program

  18. Diablo Canyon refueling outage program

    International Nuclear Information System (INIS)

    McLane, W.B.; Irving, T.L.

    1991-01-01

    Management of outages has become one of the most talked about subjects in the nuclear power industry in the past several years. Many utilities do not perform refueling outages very well or in the past have had some outages that they would not like to repeat and in some cases do not even like to think about. With the growing cost of energy and the demands placed on utilities to improve capacity factors, it is very easy for management to focus on shortening refueling outage durations as a prime objective in improving overall corporate performance. So it is with Pacific Gas and Electric Company and the Diablo Canyon power plant. A review of their refueling outage performance reflects a utility that is responding to the nuclear industry's call for improved outage performance

  19. Diablo Canyon ECCS enhancements

    International Nuclear Information System (INIS)

    Lin, A.; Lee, T.P.; Walter, L.E.

    2004-01-01

    Diablo Canyon Power Plant (DCPP) operated by Pacific Gas and Electric Co. (PG and E) is a Westinghouse designed four loop plant. In recent years, several issues were identified regarding the compliance of the Emergency Core Cooling System (ECCS) surveillance tests to the ECCS analyses assumptions. These concerns are related mostly to the High Head Safety Injection (HHSI) and the Intermediate Head Safety Injection (IHSI) systems where the injection line throttle valves are adjusted during outage surveillance testing to ensure compliance with the Technical Specifications (TS). To resolve all of the identified issues PG and E performed an ECCS reanalysis and upgraded the ECCS surveillance test program and also had Westinghouse perform a containment reanalysis using their latest model. As a result of these plant specific enhancement efforts, DCPP widened the operating window for TS surveillance testing, lowered the ECCS pumps' acceptance performance curves, and re-gained Peak Clad Temperature (PCT) and containment peak pressure margins. These enhancements are generically applicable to other plants and are addressed in this paper. (author)

  20. Computer-aided performance monitoring program at Diablo Canyon

    International Nuclear Information System (INIS)

    Nelson, T.; Glynn, R. III; Kessler, T.C.

    1992-01-01

    This paper describes the thermal performance monitoring program at Pacific Gas ampersand Electric Company's (PG ampersand E's) Diablo Canyon Nuclear Power Plant. The plant performance monitoring program at Diablo Canyon uses the THERMAC performance monitoring and analysis computer software provided by Expert-EASE Systems. THERMAC is used to collect performance data from the plant process computers, condition that data to adjust for measurement errors and missing data points, evaluate cycle and component-level performance, archive the data for trend analysis and generate performance reports. The current status of the program is that, after a fair amount of open-quotes tuningclose quotes of the basic open-quotes thermal kitclose quotes models provided with the initial THERMAC installation, we have successfully baselined both units to cycle isolation test data from previous reload cycles. Over the course of the past few months, we have accumulated enough data to generate meaningful performance trends and, as a result, have been able to use THERMAC to track a condenser fouling problem that was costing enough megawatts to attract corporate-level attention. Trends from THERMAC clearly related the megawatt loss to a steadily degrading condenser cleanliness factor and verified the subsequent gain in megawatts after the condenser was cleaned. In the future, we expect to rebaseline THERMAC to a beginning of cycle (BOC) data set and to use the program to help track feedwater nozzle fouling

  1. Loss of residual heat removal system: Diablo Canyon, Unit 2, April 10, 1987

    International Nuclear Information System (INIS)

    1987-06-01

    This report presents the findings of an NRC Augmented Inspection Team (AIT) investigation into the circumstances associated with the loss of residual heat removal (RHR) system capability for a period of approximately one and one-half hours at the Diablo Canyon, Unit 2 reactor facility on April 10, 1987. This event occurred while the Diablo Canyon, Unit 2, a pressurized water reactor, was shutdown with the reactor coolant system (RCS) water level drained to approximately mid-level of the hot leg piping. The reactor containment building equipment hatch was removed at the time of the event, and plant personnel were in the process of removing the primary side manways to gain access into the steam generator channel head areas. Thus, two fission product barriers were breached throughout the event. The RCS temperature increased from approximately 87 0 F to bulk boiling conditions without RCS temperature indication available to the plant operators. The RCS was subsequently pressurized to approximately 7 to 10 psig. The NRC AIT members concluded that the Diablo Canyon, Unit 2 plant was, at the time of the event, in a condition not previously analyzed by the NRC staff. The AIT findings from this event appear significant and generic to other pressurized water reactor facilities licensed by the NRC

  2. Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 30

    International Nuclear Information System (INIS)

    1985-04-01

    Supplement 30 to the Safety Evaluation Report for the application by the Pacific Gas and Electric Company (PG and E) to operate the Diablo Canyon Nuclear Power Plant - Unit 2 (Docket No. 50-323) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. SSER 30 reports on the staff's technical review and evaluation of the design and analysis of Unit 2 piping systems and pipe supports. The staff effort includes an evaluation of PG and E's treatment of issues raised during the Unit 1 design verification, actions resulting from low power License Condition 2.C.(11) in the Unit 1 low power license DPR-76 and the Unit 2 applicability and resolution of certain allegations related to piping and supports

  3. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    International Nuclear Information System (INIS)

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-01-01

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation or neutrino oscillation by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5 percent respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock

  4. Geologic Investigation of a Potential Site for a Next-Generation Reactor Neutrino Oscillation Experiment -- Diablo Canyon, San Luis Obispo County, CA

    International Nuclear Information System (INIS)

    Onishi, Celia Tiemi; Dobson, Patrick; Nakagawa, Seiji; Glaser, Steven; Galic, Dom

    2004-01-01

    This report provides information on the geology and selected physical and mechanical properties of surface rocks collected at Diablo Canyon, San Luis Obispo County, California as part of the design and engineering studies towards a future reactor neutrino oscillation experiment. The main objective of this neutrino project is to study the process of neutrino flavor transformation--or neutrino oscillation--by measuring neutrinos produced in the fission reactions of a nuclear power plant. Diablo Canyon was selected as a candidate site because it allows the detectors to be situated underground in a tunnel close to the source of neutrinos (i.e., at a distance of several hundred meters from the nuclear power plant) while having suitable topography for shielding against cosmic rays. The detectors have to be located underground to minimize the cosmic ray-related background noise that can mimic the signal of reactor neutrino interactions in the detector. Three Pliocene-Miocene marine sedimentary units dominate the geology of Diablo Canyon: the Pismo Formation, the Monterey Formation, and the Obispo Formation. The area is tectonically active, located east of the active Hosgri Fault and in the southern limb of the northwest trending Pismo Syncline. Most of the potential tunnel for the neutrino detector lies within the Obispo Formation. Review of previous geologic studies, observations from a field visit, and selected physical and mechanical properties of rock samples collected from the site provided baseline geological information used in developing a preliminary estimate for tunneling construction cost. Gamma-ray spectrometric results indicate low levels of radioactivity for uranium, thorium, and potassium. Grain density, bulk density, and porosity values for these rock samples range from 2.37 to 2.86 g/cc, 1.41 to 2.57 g/cc, and 1.94 to 68.5% respectively. Point load, unconfined compressive strength, and ultrasonic velocity tests were conducted to determine rock mechanical

  5. Safety evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Suppl. 22

    International Nuclear Information System (INIS)

    1984-03-01

    Supplement 22 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plants, Unit 1 and 2 (Docket Nos. 50-275 and 50-323), has been prepared jointly by the Office of Nuclear Reactor Regulation and the Region V Office of the US Nuclear Regulatory Commission. This supplement provides the criteria that were used by the staff to determine which of the allegations that have been evaluated and must be resolved prior to Unit 1 achieving criticality and operating at power level up to 5 percent of rated power (i.e., low power operation). The supplement also reports on the status of the staff's investigation, inspection and evaluation of 219 allegations or concerns that have been identified to the NRC as of March 9, 1984, excluding those recently received under 10 CFR 2.206 petitions

  6. Auxiliary feedwater system risk-based inspection guide for the Diablo Canyon Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gore, B.F.; Vo, T.V.; Harrison, D.G.

    1990-08-01

    This document presents a compilation of auxiliary feedwater (AFW) system failure information which has been screened for risk significance in terms of failure frequency and degradation of system performance. It is a risk-prioritized listing of failure events and their causes that are significant enough to warrant consideration in inspection planning at Diablo Canyon. This information is presented to provide inspectors with increased resources for inspection planning at Diablo Canyon. The risk importance of various component failure modes was identified by analysis of the results of probabilistic risk assessments (PRAs) for many pressurized water reactors (PWRs). However, the component failure categories identified in PRAs are rather broad, because the failure data used in the PRAs is an aggregate of many individual failures having a variety of root causes. In order to help inspectors to focus on specific aspects of component operation, maintenance and design which might cause these failures, an extensive review of component failure information was performed to identify and rank the root causes of these component failures. Both Diablo Canyon and industry-wide failure information was analyzed. Failure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorized as common cause failures, human errors, design problems, or component failures. This information permits an inspector to concentrate on components important to the prevention of core damage. Other components which perform essential functions, but which are not included because of high reliability or redundancy, must also be addressed to ensure that degradation does not increase their failure probabilities, and hence their risk importances. 23 refs., 1 fig., 1 tab

  7. Diablo Canyon plant information management system and integrated communication system

    International Nuclear Information System (INIS)

    Stanley, J.W.; Groff, C.

    1990-01-01

    The implementation of a comprehensive maintenance system called the plant information management system (PIMS) at the Diablo Canyon plant, together with its associated integrated communication system (ICS), is widely regarded as the most comprehensive undertaking of its kind in the nuclear industry. This paper provides an overview of the program at Diablo Canyon, an evaluation of system benefits, and highlights the future course of PIMS

  8. Diablo Canyon plant information management system and integrated communication system

    Energy Technology Data Exchange (ETDEWEB)

    Stanley, J.W.; Groff, C.

    1990-06-01

    The implementation of a comprehensive maintenance system called the plant information management system (PIMS) at the Diablo Canyon plant, together with its associated integrated communication system (ICS), is widely regarded as the most comprehensive undertaking of its kind in the nuclear industry. This paper provides an overview of the program at Diablo Canyon, an evaluation of system benefits, and highlights the future course of PIMS.

  9. 33 CFR 165.1155 - Security Zone; Diablo Canyon Nuclear Power Plant, Avila Beach, California.

    Science.gov (United States)

    2010-07-01

    ... Nuclear Power Plant, Avila Beach, California. 165.1155 Section 165.1155 Navigation and Navigable Waters... Coast Guard District § 165.1155 Security Zone; Diablo Canyon Nuclear Power Plant, Avila Beach... surface to bottom, within a 2,000 yard radius of Diablo Canyon Nuclear Power Plant centered at position 35...

  10. Diablo Canyon: the challenges of the Californian nuclear

    International Nuclear Information System (INIS)

    Avrin, Anne-Perrine; Zweibaum, Nicolas

    2014-01-01

    This article addresses the Californian nuclear plant of Diablo Canyon which, due to the shutting down of the two reactors of another plant, could be the last nuclear plant to operate in California as this state voted an interdiction of construction of any new nuclear plant until a permanent solution has been found for radioactive waste storage. It is outlined that this nuclear plant is the most controlled nuclear plant throughout the USA, presents an unchallenged security level despite it is built at the vicinity of seismic fault lines. It outlines that the problem of waste storage is far from being solved in California, and discusses the future role of nuclear energy in California within the context of global warming

  11. Aerial radiological survey of the Diablo Canyon Nuclear Power Plant and surrounding area, Diablo Canyon, California. Date of survey: September-October 1984

    International Nuclear Information System (INIS)

    1985-03-01

    An aerial radiological survey was conducted over the area surrounding the Diablo Canyon Nuclear Power Plant in Diablo Canyon, California. The survey was conducted between 20 September and 3 October 1984. A series of flight lines parallel to the coastline were flown at an altitude of 91 meters (300 feet) and were spaced 152 meters (500 feet) apart. The survey covered an area of 250 square kilometers (100 square miles). The resulting background exposure rates over the survey area ranged from 5 to 21 microroentgens per hour (μR/h). The reported exposure rate values include an estimated cosmics ray contribution of 3.6 μR/h. Soil samples were also collected at several locations within the survey areas and analyzed in the laboratory for isotopic composition. The results of the survey showed only the presence of naturally occurring background radiation. No man-made radioactivity was detected. 4 refs., 4 figs., 4 tabs

  12. Seismic qualification of the rotary relay for use in the Trojan and Diablo Canyon Auxiliary Safeguards Cabinets

    International Nuclear Information System (INIS)

    Riggio, M.D.; Jarecki, S.J.

    1977-10-01

    This report presents the results of the analysis performed for the seismic qualification of the rotary relay for use in the Trojan and Diablo Canyon Auxiliary Safeguards Cabinets. A finite element model of the cabinet was developed from seismic test results. This model was analytically subjected to a simulated 3D floor acceleration time history that enveloped, simultaneously, the Trojan and the June 1969 Diablo Canyon Safe Shutdown Earthquake requirements. The dynamic response of the cabinet at the mounting location of the rotary relays was determined. The calculated acceleration time histories were converted to response spectra and these response spectra were compared to the test response spectra successfully achieved during the rotary relay seismic qualification tests. It was found that the dynamic motion levels at the rotary relays, when mounted in the Trojan or Diablo Canyon Auxiliary Safeguards Cabinets, do not exceed the levels for which they were previously seismically qualified by tests. Consequently, the rotary relays are seismically qualified for use in the Trojan or Diablo Canyon Auxiliary Safeguards Cabinets

  13. Socio-economic impacts of nuclear generating stations: Diablo Canyon case study

    International Nuclear Information System (INIS)

    Pijawka, K.D.; Yaquinto, G.

    1982-07-01

    This report documents a case study of the socioeconomic impacts of the construction and operation of the Diablo Canyon nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980 to 1981. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study area during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  14. Pilot RCM application to the Diablo Canyon main stream system

    International Nuclear Information System (INIS)

    Groff, C.R.; Beckham, P.E.; Bych, K.H.

    1988-01-01

    In 1986 Pacific Gas ampersand Electric Company (PG ampersand E) became extremely interested in reliability-centered maintenance (RCM) after the initial review of two successful Electric Power Research Institute sponsored projects. RCM was visualized as a methodology to common sensitize the burgeoning preventive maintenance (PM) program at the Diablo Canyon plant. RCM could further the uses of predictive and condition-monitoring techniques, as well as eliminate maintenance on components whose failures were noncritical. An extensive review of maintenance and operation experience data, in conjunction with plant staff recommendations and a prioritization according to maintenance expenditures and operational/safety significance, produced the selected system: the turbine main steam supply system (main steam). The pilot project segmented the main steam system into eight subsystems to aid in analysis: (a) main steam isolation valves, (b) auxiliary feedwater pump turbine, (c) overpressure protection (steam dump), (d) main feedwater pump turbines, (e) main steam, (f) main turbine, (g) steam blowdown, and (h) moisture separator reheaters. System analysis activities, including the preparation of functional failure analyses, failure modes and effects analyses, and logic model analyses, were conducted in parallel with corrective and preventive maintenance data-gathering activities to maximize project team personnel participation during the project. Results and lessons learned are summarized

  15. Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Enos, David George [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-07-01

    Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

  16. Socioeconomic impacts of nuclear generating stations - Diablo Canyon case study. Technical report for 1 Oct 78-4 Jan 82

    International Nuclear Information System (INIS)

    Pijawka, K.D.; Yaquinto, G.

    1982-07-01

    This report documents a case study of the socioeconomic impacts of the construction and operation of the Diablo Canyon nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlemeent patterns and housing, local government and public services, social structure, and public response in the study during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

  17. A computerized assessment and response system for radiological emergency at Diablo Canyon Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shih, C.C.; Thuillier, R.H.

    1984-01-01

    The U.S. Nuclear Regulatory Commission requires that nuclear power plants provide for rapid assessment and response in the event of a radiological emergency. At the Diablo Canyon Nuclear Power Plant, Pacific Gas and Electric Company uses a system of linked central minicomputer, satellite desktop computers and microprocessors to provide decision makers with timely and pertinent information in emergency situations. The system provides for data acquisition and microprocessing at meteorological and radiological monitoring sites. Current estimates or projections of offsite dose commitment are made in real-time by a dispersion/dose calculation model. Computerized dissemination of data and calculational results to decision makers at the government and utility levels is also available. The basic system in use is a commercially available Emergency Assessment and Response System (EARS). This generic system has been modified in-house to meet requirements specific to emergency situations at the plant. Distinctive features of the modification program includes: a highly professional man-machine interaction; consideration of site-specific factors; simulation of environmental radiology for development of drill scenarios; and concise, pertinent reports as input to decision making

  18. Outage risk reduction at Diablo Canyon

    International Nuclear Information System (INIS)

    Burnett, Tobias W.T.; Eugene Newman, C.

    2004-01-01

    A formal risk reduction program was conducted at the Diablo Canyon Nuclear Generating plant as part of EPRI's Outage Risk Assessment and Management Program. The program began with a probabilistic and deterministic assessment of the frequency of core coolant boiling and core uncovery during shutdown operations. This step identified important contributors to risk, periods of high vulnerability, and potential mechanisms for reducing risk. Next, recovery strategies were evaluated and procedures, training, and outage schedules modified. Twelve risk reduction enhancements were developed and implemented. These enhancements and their impact are described in this paper. These enhancements reduced the calculated risk of core uncovery by about a factor of four for a refueling outage without lengthening the outage schedule; increased the outage efficiency, contributing to completing 11 days ahead of schedule; and helped to earn the highest achievable SALP rating from the NRC. (author)

  19. Safety-evaluation report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323). Supplement No. 16

    International Nuclear Information System (INIS)

    1983-08-01

    This supplement addresses certain allegations received by the staff concerning safety deficiencies at the Diablo Canyon project, specifically those relative to design aspects of the component cooling water system (CCWS), including its seismic classification, single failure considerations, and its heat removal capabilities. It is based on the staff's re-review of the Diablo Canyon Final Safety Analysis Report (FSAR); on a site walkdown conducted on January 12, 1983; on information obtained during meetings with the applicant dated March 15, 18, and 25, April 4, 7, and 15, and May 3 and 18, 1983

  20. Technical evaluation report TMI action: NUREG-0737 (II.D.1) relief and safety valve testing,. Diablo Canyon Units 1 and 2 (Docket Nos. 50-275, 50-323)

    International Nuclear Information System (INIS)

    Miller, G.K.; Magleby, H.L.; Nalezny, C.L.

    1984-07-01

    Light water reactor operators have experienced a number of occurrences of improper performance of safety and relief valves installed in their primary coolant systems. Because of this, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Status Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient and accident conditions. This report provides the results of the review of these programs and their results by the Nuclear Regulatory Commission (NRC) and their consultant, EG and G Idaho, Inc. Specifically, this report has examined the response of the Licensee for Diablo Canyon Units 1 and 2, to the requirements of NUREG-0578 and NUREG-0737 and finds that the Licensee has provided an acceptable response, reconfirming that the General Design Criteria 14, 15 and 30 of Appendix A to 10 CRF 50 have been met. 18 refs

  1. Public response to the Diablo Canyon Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Pijawka, K D [Arizona State Univ., Tempe (USA)

    1982-08-01

    We examine the nature of the public response to the Diablo Canyon Nuclear Generating Station located in San Luis Obispo, California, from the early 1960s to the present. Four distinct phases of public intervention were discerned, based on change in both plant-related issues and in the nature of the antinuclear constituencies in the region. The level of public concern varied both geographically and temporally and is related to the area's social structure, environmental predispositions, and distribution of plant-related economic benefits. External events, such as the prolonged debate over the risk assessment of the seismic hazard and the Three Mile Island accident were found to be important factors in explaining variation in public concern and political response.

  2. Public response to the Diablo Canyon Nuclear Generating Station

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-01-01

    The authors examine the nature of the public response to the Diablo Canyon Nuclear Generating Station located in San Luis Obispo, California, from the early 1960s to the present. Four distinct phases of public intervention were discerned, based on change in both plant-related issues and in the nature of the antinuclear constituencies in the region. The level of public concern varied both geographically and temporally and is related to the area's social structure, environmental predispositions, and distribution of plant-related economic benefits. External events, such as the prolonged debate over the risk assessment of the seismic hazard and the Three Mile Island accident were found to be important factors in explaining variation in public concern and political response

  3. Public response to the Diablo Canyon Nuclear Generating Station

    International Nuclear Information System (INIS)

    Pijawka, K.D.

    1982-01-01

    We examine the nature of the public response to the Diablo Canyon Nuclear Generating Station located in San Luis Obispo, California, from the early 1960s to the present. Four distinct phases of public intervention were discerned, based on change in both plant-related issues and in the nature of the antinuclear constituencies in the region. The level of public concern varied both geographically and temporally and is related to the area's social structure, environmental predispositions, and distribution of plant-related economic benefits. External events, such as the prolonged debate over the risk assessment of the seismic hazard and the Three Mile Island accident were found to be important factors in explaining variation in public concern and political response. (author)

  4. Specifications, tests, and installation of wires and cables for the Diablo Canyon Nuclear Power Project

    International Nuclear Information System (INIS)

    Dan, F.J.

    1977-01-01

    The process of selecting wires and cables for the Diablo Canyon Nuclear Power Project is described. The criteria for the fire and environmental tests, the basis for the specifications, and the reasons for the final choice and acceptance are outlined. A short section is dedicated to the installation of cables in raceways with reference to separation and color coding. Also covered are the selection and testing of fire stops and the selection of seismic supports

  5. Θ13 Neutrino Experiment at the Diablo Canyon Power Plant. LBNL Engineering Summary Report

    International Nuclear Information System (INIS)

    Oshatz, Daryl

    2004-01-01

    This summary document describes the results of conceptual design and cost estimates performed by LBNL Engineering staff between October 10, 2003 and March 12, 2004 for the proposed θ 13 neutrino experiment at the Diablo Canyon Power Plant (DCPP). This document focuses on the detector room design concept and mechanical engineering issues associated with the neutrino detector structures. Every effort has been made not to duplicate information contained in the last LBNL Engineering Summary Report dated October 10, 2003. Only new or updated information is included in this document

  6. Nuclear Regulatory Commission Issuances, September 1981

    International Nuclear Information System (INIS)

    1981-01-01

    Contents include: Issuances of the Nuclear Regulatory Commission--Commonwealth Edison Company (Dresden Nuclear Power Station, Unit 1), Consolidated Edison Company of New York (Indian Point, Unit 2), Metropolitan Edison Company, et al. (Three Mile Island Nuclear Station, Unit 1), Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Power Authority of the State of New York (Indian Point, Unit 3), Texas Utilities Generating Company, et al. (Comanche Peak Steam Electric Station, Units 1 and 2); Issuances of Atomic Safety and Licensing Appeal Boards--Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Philadelphia Electric Company, et al. (Peach Bottom Atomic Power Statin, Units 2 and 3), Metropolitan Edison Company, et al. (Three Mile Island Nuclear Statin, Unit No. 2), Public Service Electric and Gas Company (Hope Creek Generating Station, Units 1 and 2), The Toledo Edison Company, et al. (Davis-Besse Nuclear Power Station, Units 2 and 3); Issuances of the Atomic Safety Licensing Boards--Cleveland Electric Illuminating Company, et al. (Perry Nuclear Power Plant, Units 1 and 2), Commonwealth Edison Company (Dresden Station, Units 2 and 3), Houston Lighting and Power Company (Allens Creek Nuclear Generating Station, Unit 1), Southern California Edison Company, et al. (San Onofre Nuclear Generating Station, Units 2 and 3), Texas Utilities Generating Company, et al. (Comanche Peak Steam Electric Station, Units 1 and 2), Texas Utilities Generating Company, et al

  7. Diablo Canyon internal events PRA [Probabilistic Risk Assessment] review: Methodology and findings

    International Nuclear Information System (INIS)

    Fitzpatrick, R.G.; Bozoki, G.; Sabek, M.

    1990-01-01

    The review of the Diablo Canyon Probabilistic Risk Assessment (DCRPA) incorporated some new and innovative approaches. These were necessitated by the unprecedented size, scope and level of detail of the DCRPA, which was submitted to the NRC for licensing purposes. This paper outlines the elements of the internal events portion of the review citing selected findings to illustrate the various approaches employed. The paper also provides a description of the extensive and comprehensive importance analysis applied by BNL to the DCRPA model. Importance calculations included: top event/function level; individual split fractions; pair importances between frontline-support and support-support systems; system importance by initiator; and others. The paper concludes with a brief discussion of the effectiveness of the applied methodology. 3 refs., 5 tabs

  8. Achieving quality excellence at the Diablo Canyon Nuclear Power Plant

    International Nuclear Information System (INIS)

    Skidmore, S.M.; Taggart, D.A.

    1988-01-01

    Quality assurance methods at the Diablo Canyon plant were transformed from the then typical industry practices that often alienated professional and technical people, as well as craftsmen and their foremen, to a cooperative method that allowed plant personnel to work together as a team. It has created an attitude to do it right the first time. The roles of quality professionals were expanded to include teaching and coaching to facilitate enhanced communication between and within functional organizations. This included regular presentations to managers and line personnel in an informal group participative atmosphere. These presentations have become widely known at the plant as quality awareness tailboard sessions. These presentations are intended to increase personnel sensitivity to the subject of quality and quality management. Economic achievement of excellence in quality is essential to remain competitive in today's marketplace. The proactive team-oriented approach of quality assurance achieves the bottom line of high quality with concurrently enhanced productivity and cost-effectiveness

  9. Ion exchange media testing for processing recyclable and nonrecyclable liquids at Diablo Canyon Power Plant

    International Nuclear Information System (INIS)

    James, K.L.; Miller, C.C.

    1989-01-01

    This paper reports on several ion exchange materials tested for processing nonrecyclable and recyclable liquid wastes at Diablo Canyon Power Plant. These ion exchange materials include inorganic Durasil media, natural and synthetic zeolites, and various organic resins. Additional tests were performed using a polyelectrolyte pretreatment technique to enhance processing of liquid wastes by ion exchange. A 9:1 ratio of cation to anion resin, consisting of IRN-77 and Sybron A-642 was effective in decontaminating cesium and cobalt radionuclides for low conductivity nonrecyclable liquids. A mixture of zeolite and Durasil media was most effective in removing cesium and cobalt from nonrecyclable high conductivity liquids. The experimental Dow resins achieved the best results in decontaminating recyclable liquids and minimized the effluent levels of chlorides, sulfates, and silica

  10. Application of Landsat Thematic Mapper data for coastal thermal plume analysis at Diablo Canyon

    Science.gov (United States)

    Gibbons, D. E.; Wukelic, G. E.; Leighton, J. P.; Doyle, M. J.

    1989-01-01

    The possibility of using Landsat Thematic Mapper (TM) thermal data to derive absolute temperature distributions in coastal waters that receive cooling effluent from a power plant is demonstrated. Landsat TM band 6 (thermal) data acquired on June 18, 1986, for the Diablo Canyon power plant in California were compared to ground truth temperatures measured at the same time. Higher-resolution band 5 (reflectance) data were used to locate power plant discharge and intake positions and identify locations of thermal pixels containing only water, no land. Local radiosonde measurements, used in LOWTRAN 6 adjustments for atmospheric effects, produced corrected ocean surface radiances that, when converted to temperatures, gave values within approximately 0.6 C of ground truth. A contour plot was produced that compared power plant plume temperatures with those of the ocean and coastal environment. It is concluded that Landsat can provide good estimates of absolute temperatures of the coastal power plant thermal plume. Moreover, quantitative information on ambient ocean surface temperature conditions (e.g., upwelling) may enhance interpretation of numerical model prediction.

  11. Nuclear Regulatory Commission issuances. Volume 40, Number 5

    International Nuclear Information System (INIS)

    1994-11-01

    This book contains issuances of the Atomic Safety and Licensing Boards for November 1994. The issuances include Cameo Diagnostic Centre, Inc. byproduct material license; Georgia Power Company license amendment, transfer to Southern Nuclear for Vogtle Electric Generating Plant, units 1 and 2; Indiana Regional Cancer Center, order modifying and suspending byproduct material license; Louisiana Energy Services, special nuclear material license; Pacific Gas and Electric Company, construction period recovery, facility operating license, Diablo Canyon Nuclear Power plant; and Sequoyah Fuels Corporation, source materials license

  12. 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

  13. Three Years of Experience of Wet Gas Allocation on Canyon Express

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Aditya; Hall, James; Letton, Winsor

    2005-07-01

    In September 2002, production was begun from the three fields that together form the Canyon Express System- King's Peak, Aconcagua, and Camden Hills. The 9 wells from these fields are connected to a pair of 12-inch flow lines carrying the commingled wet gas a distance of approximately 92 kilometers back to the Canyon Station platform for processing. At the 21st NSFMW in October 2003, an initial report was given on the status of Wet Gas Allocation for the Canyon Express project. As discussed in that paper, dual-differential, subsea wet gas meters were chosen for the task of allocating gas and liquids back to individual wells. However, since the gas from all three fields was very dry (Lockhart-Martinelli parameter less than 0.01) and because the operating pressures were quite high (250 bar), application of the dual-differential function of the meters yielded errors in both liquid and gas flow rates. Furthermore, as these problems were being uncovered, scale was beginning to collect inside some of the meters. Taken together, these problems produced system imbalances as great as 20%. To address the problems, one of the individual flow metering elements within each wet gas meter was chosen as the allocation meter, operating as a single-phase gas meter. After three years of operation of the Canyon Express Project, considerable experience has been accumulated. Since at the time it held the record for deep water hydrocarbon production, application of the technologies discussed here were challenging and required considerable flexibility. It is believed that the Canyon Express experiences will benefit future deep water flow metering projects. The knowledge acquired thus far is surveyed and summarized. The emphasis is on the technical aspects. (tk)

  14. 10 blows that stopped nuclear power

    International Nuclear Information System (INIS)

    Komanoff, C.

    1991-01-01

    The author describes these 10 blows in chronological order, 1973 through 1981, namely: (1) Arab Oil Embargo; (2) India Explodes a Bomb; (3) NRC replaces AEC; (4) Fire at Browns Ferry; (5) General Electric and NRC Engineers switch Sides; (6) Amory Lovins Recasts the Energy Debate; (7) The Seabrook Occupation; (8) The Three Mile Island Accident; (9) Federal Reserve Tightens the Money Supply; and (1) Pacific Gas and Electric Co. Gets it Backwards at Diablo Canyon. he stops there, not including the Washington Public Power Supply fiasco and the Chernobyl disaster, feeling nuclear expansion was essentially foreclosed without them. Further, he feels nuclear power seems fated to be forever at the mercy of forces beyond its control

  15. Use of action requests to control communications

    International Nuclear Information System (INIS)

    Brady, J.

    1988-01-01

    This paper discusses the Plant Information Management System (PIMS) that is implemented at Pacific Gas and Electric Company's (PG and E) Diablo Canyon Power Plant (DCPP). PIMS is implemented on IBM mainframes located at the plant, is on-line and interactive, and is accessed via a computer communication system that supports more than 450 IBM 3270 PC workstations. This paper discusses the role of the ACTION REQUEST module of PIMS and how it is used to control plant sensitive communications. The ACTION REQUEST module of PIMS can be accessed from any workstation and during the first year of Commercial Operation of DCPP replaced numerous and redundant forms of manual communication mechanisms. Also in this first year, users at the plant generated approximately 25,000 Action Requests which were controlled through review and approval cycles by PIMS. Each organization assigned action were immediately notified of their responsibilities so that action could be taken in a timely manner. The Diablo Canyon Power Plant broke Westinghouse world-wide operating records for the first year of operation (over 90% availability) due to a well built and reliable plant and due to a responsive Operations organization, which was well informed and controlled

  16. Component fragility research program

    International Nuclear Information System (INIS)

    Tsai, N.C.; Mochizuki, G.L.; Holman, G.S.

    1989-11-01

    To demonstrate how ''high-level'' qualification test data can be used to estimate the ultimate seismic capacity of nuclear power plant equipment, we assessed in detail various electrical components tested by the Pacific Gas ampersand Electric Company for its Diablo Canyon plant. As part of our Phase I Component Fragility Research Program, we evaluated seismic fragility for five Diablo Canyon components: medium-voltage (4kV) switchgear; safeguard relay board; emergency light battery pack; potential transformer; and station battery and racks. This report discusses our Phase II fragility evaluation of a single Westinghouse Type W motor control center column, a fan cooler motor controller, and three local starters at the Diablo Canyon nuclear power plant. These components were seismically qualified by means of biaxial random motion tests on a shaker table, and the test response spectra formed the basis for the estimate of the seismic capacity of the components. The seismic capacity of each component is referenced to the zero period acceleration (ZPA) and, in our Phase II study only, to the average spectral acceleration (ASA) of the motion at its base. For the motor control center, the seismic capacity was compared to the capacity of a Westinghouse Five-Star MCC subjected to actual fragility tests by LLNL during the Phase I Component Fragility Research Program, and to generic capacities developed by the Brookhaven National Laboratory for motor control center. Except for the medium-voltage switchgear, all of the components considered in both our Phase I and Phase II evaluations were qualified in their standard commercial configurations or with only relatively minor modifications such as top bracing of cabinets. 8 refs., 67 figs., 7 tabs

  17. 9th Pacific Basin Nuclear Conference. Nuclear energy, science and technology - Pacific partnership. Proceedings Volume 1

    International Nuclear Information System (INIS)

    1994-04-01

    The theme of the 9th Pacific Basin Nuclear conference held in Sydney from 1-6 May 1994, embraced the use of the atom in energy production and in science and technology. The focus was on selected topics of current and ongoing interest to countries around the Pacific Basin. The two-volume proceedings include both invited and contributed papers. They have been indexed separately. This document, Volume 1 covers the following topics: Pacific partnership; perspectives on nuclear energy, science and technology in Pacific Basin countries; nuclear energy and sustainable development; economics of the power reactors; new power reactor projects; power reactor technology; advanced reactors; radioisotope and radiation technology; biomedical applications

  18. 75 FR 54920 - In the Matter of Pacific Gas & Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2...

    Science.gov (United States)

    2010-09-09

    ...: Contention EC-1: PG&E's Severe Accident Mitigation Alternatives (``SAMA'') analysis fails to satisfy 40 CFR... it does not address the airborne environmental impacts of a spent fuel pool accident caused by an...

  19. High Smac/DIABLO expression is associated with early local recurrence of cervical cancer

    International Nuclear Information System (INIS)

    Arellano-Llamas, Abril; Garcia, Francisco J; Perez, Delia; Cantu, David; Espinosa, Magali; De la Garza, Jaime G; Maldonado, Vilma; Melendez-Zajgla, Jorge

    2006-01-01

    In a recent pilot report, we showed that Smac/DIABLO mRNA is expressed de novo in a subset of cervical cancer patients. We have now expanded this study and analyzed Smac/DIABLO expression in the primary lesions in 109 cervical cancer patients. We used immunohistochemistry of formalin-fixed, paraffin-embedded tissue sections to analyze Smac/DIABLO expression in the 109 primary lesions. Seventy-eight samples corresponded to epidermoid cervical cancer and 31 to cervical adenocarcinoma. The median follow up was 46.86 months (range 10–186). Smac/DIABLO was expressed in more adenocarcinoma samples than squamous tumours (71% vs 50%; p = 0.037). Among the pathological variables, a positive correlation was found between Smac/DIABLO immunoreactivity and microvascular density, a marker for angiogenesis (p = 0.04). Most importantly, Smac/DIABLO immunoreactivity was associated with a higher rate of local recurrence in squamous cell carcinoma (p = 0.002, log rank test). No association was found between Smac/DIABLO and survival rates. Smac/DIABLO expression is a potential marker for local recurrence in cervical squamous cell carcinoma patients

  20. Impact of oil and gas infrastructure development in La Manga Canyon, NM

    Science.gov (United States)

    La Manga Canyon is a small watershed (~20km2) in the San Juan Basin that has historically been developed for natural gas and recently for coal bed methane. Since gas production began in the 1940s, an extensive network of dirt roads have transected the watershed, providing access to well sites. There...

  1. NRC licensing of Diablo Canyon. Hearing before the Subcommittee on Energy Conservation and Power of the Committee on Energy and Commerce, House of Representatives, Ninety-Ninth Congress, First Session, July 10, 1985

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    The members of the Nuclear Regulatory Commission (NRC) and two California members of Congress testified at a hearing to examine the decision-making capacity and the integrity of the NRC. The specific issue was whether the issuance of an operating license to the Diablo Canyon nuclear plant violated the Atomic Energy Act in the area of safety. The transcripts of closed meetings of the NRC and the expression of concern by California congressmen and a member of the NRC about the Commission's decision-making process prompted the hearings. Specific concerns were the possible avoidance of a public hearing on emergency plans in the event of an earthquake and to avoid the costs of a hearing and the quality of information used as a basis for the decision. Each member of the Commission testified in response to these concerns. The California representatives noted that Commissioners did not follow legal advice, and that they relied upon second-hand information. Other material and documents submitted for the record follows the testimony

  2. 75 FR 12315 - Pacific Gas and Electric Company; Diablo Canyon Independent Spent Fuel Storage Installation...

    Science.gov (United States)

    2010-03-15

    ... be filed in accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing... the procedural requirements of E-Filing, at least ten (10) days prior to the filing deadline, the... the NRC in accordance with the E-Filing rule, the participant must file the document using the NRC's...

  3. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    1974-01-01

    The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

  4. NRC Information No. 88-04, Supplement 1: Inadequate qualification and documentation of fire barrier penetration seals

    International Nuclear Information System (INIS)

    Rossi, C.E.

    1992-01-01

    Since August 9, 1979, Diablo Canyon has experienced four fires in which the penetration seal material ignited and burned. On March 25, 1986, Toledo Edison reported that the silicone foam sealant in a fire barrier penetration in the main steam line room of Davis-Besse appeared to have baked and pulled away from the pipe. This penetration was filled with low-density silicone foam, which is rated for a maximum temperature of 425 F. The main steam lines normally operate near 600 F. The licensee's corrective actions were to stuff ceramic fiber (Kaowool) into the seal for increased fire resistance and to expand the existing fire watch. The silicone foam seal material in the diesel generator exhaust pipe penetrations at Diablo Canyon was used as replacement material for the combustible foam plastic-type seals, which were involved in the 1975 fire at the Browns Ferry nuclear power plant. In addition to diesel generator exhaust pipe penetrations, some main steamline penetrations may be sealed with this material. The material is apparently designed to withstand maximum ambient temperatures of 400 F. and temporary exposure to 500 F., but not long-term exposure to higher temperatures. Although the measurement at Diablo Canyon in 1982 indicated a pipe temperature of about 600 F., average pipe temperatures are probably much higher. The licensee for Diablo Canyon determined that only six penetrations (all associated with diesel generator exhaust pipes) potentially exposed the silicone foam penetration seal material to high temperatures. For these penetration openings, the licensee plans to install a penetration seal material designed to withstand long-term exposure to high temperatures

  5. Nuclear Regulatory Commission Issuances, May 81

    International Nuclear Information System (INIS)

    1981-05-01

    Contents: Issuances of the Nuclear Regulatory Commission--Consolidated Edison Company of New York, Inc. (Indian Point, Unit No. 2), Power Authority of the State of New York (Indian Point, No. 3 Nuclear Power Plant), Pacific Gas and Electric Company (Diablo Canyon Nuclear Power Plant, Units 1 and 2), Statement of Policy on Conduct of Licensing Proceedings, Uranium Mill Licensing Requirements; Issuances of Atomic Safety and Licensing Appeal Boards--Houston Lighting and Power Company, et al. (South Texas Project, Units 1 and 2), Metropolitan Edison Company, et al. (Three Mile Island Nuclear Station, Unit No. 2), Pennsylvania Power and Light Company and Allegheny Electric Cooperative, Inc. (Susquehanna Steam Electric Station, Units 1 and 2), Philadelphia Electric Company et al. (Peach Bottom Atomic Power Station, Units 2 and 3), Public Service Electric and Gas Company (Hope Creek Generating Station, Units 1 and 2); Issuances of the Atomic Safety and Licensing Boards--Duke Power Company (William B. McGuire Nuclear Station, Units 1 and 2), Florida Light and Power Company (Turkey Point Nuclear Generating, Units 3 and 4), Illinois Power Company, et al. (Clinton Power Station, Units 1 and 2), Sacramento Municipal Utility District (Rancho Seco Nuclear Generating Station); Issuances of the Directors Denial--Commonwealth Edison Company (Byron Station, Units 1 and 2), Consolidated Edison Company of New York, Inc. (Indian Point Unit No. 2), Gulf States Utilities Company (River Bend Station Units 1 and 2), Petition to Suspend All Operating Licenses for Pressurized Water Reactors (River Bend Station Units 1 and 2), Portland General Electric Company (Trojan Nuclear

  6. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Han, K.I.

    1977-01-01

    Preliminary investigations of a heterogeneous gas core reactor (HGCR) concept suggest that this potential power reactor offers distinct advantages over other existing or conceptual reactor power plants. One of the most favorable features of the HGCR is the flexibility of the power producing system which allows it to be efficiently designed to conform to a desired optimum condition without major conceptual changes. The arrangement of bundles of moderator/coolant channels in a fissionable gas or mixture of gases makes a truly heterogeneous nuclear reactor core. It is this full heterogeneity for a gas-fueled reactor core which accounts for the novelty of the heterogeneous gas core reactor concept and leads to noted significant advantages over previous gas core systems with respect to neutron and fuel economy, power density, and heat transfer characteristics. The purpose of this work is to provide an insight into the design, operating characteristics, and safety of a heterogeneous gas core reactor system. The studies consist mainly of neutronic, energetic and kinetic analyses of the power producing and conversion systems as a preliminary assessment of the heterogeneous gas core reactor concept and basic design. The results of the conducted research indicate a high potential for the heterogeneous gas core reactor system as an electrical power generating unit (either large or small), with an overall efficiency as high as 40 to 45%. The HGCR system is found to be stable and safe, under the conditions imposed upon the analyses conducted in this work, due to the inherent safety of ann expanding gaseous fuel and the intrinsic feedback effects of the gas and water coolant

  7. Slope instabilities along the Western Andean Escarpment and the main canyons in Northern Chile

    Science.gov (United States)

    Crosta, G.; Hermanns, R. L.; Valbuzzi, E.; Dehls, J.; Yugsi Molina, F. X.; Sepulveda, S.

    2012-04-01

    The western slope of the Andes of northern Chile - southern Perù is generally subdivided from the west to the east into the morphological units of: the Coastal Cordillera, Central Depression, the Western Escarpment-Precordillera and the Western Andean Cordillera. The western escarpment and Precordillera are formed by the Azapa coarse-grained clastic formation (sandstones, conglomerates, mudstones) and the Oxaya (rhyodacitic ignimbrites) and Diablo volcanoclastic formations (Oligocene and Miocene). Important uplift has been suggested between the deposition of the Oxaya and Diablo formations. The entire area has been characterized by a long-term hyperaridity (Atacama desert), initially established between 20 and 15 Ma, and this caused a strong difference between the long term continuous uplift and low denudation rates. This long sector of the central western escarpment and Precordillera is incised by deep canyons and subparallel drainage network in the upper part. The drainage network developed in two main phases: a lower-middle Miocene phase with formation of a parallel poorly structured drainage network cutting into the Oxaya formation, and presently well preserved; the canyons have been incised in the initial topography starting around 9 Ma and up to about 3.8 Ma with subsequent refilling episodes. Valley incision (ave. rate of 0.2 mm yr-1) has been controlled by topographic uplift and less arid climate (after 7 Ma). As a consequence of these geologic and climatic settings the evolution of this area has been characterized by canyon incision and extremely large slope instabilities. These slope instabilities occur in the "interfluvial" sectors of the western escarpment and Precordillera and along the canyon flanks. Landslides affecting the preserved paleosurfaces, interested by the parallel drainage network in the Oxaya formation, involve volumes of various cubic kilometres (Lluta collapse, Latagualla Landslide) and can control the drainage network. These mega

  8. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Vakilian, M.

    1977-05-01

    The present study is the second part of a general survey of Gas Cooled Reactors (GCRs). In this part, the course of development, overall performance and present development status of High Temperature Gas Cooled Reactors (HTCRs) and advances of HTGR systems are reviewed. (author)

  9. The Pacific Rim and global natural gas

    International Nuclear Information System (INIS)

    Dreyfus, D.A.

    1993-01-01

    There is a growing interest in natural gas as a part of national or international strategies to moderate the environmental consequences of fuel use. Although the underutilized global gas resource justifies the interest, the future consumption of gas is likely to be constrained by the high capital costs of new transportation facilities to bring remote gas supplies into areas of growing energy demand. The Asian Pacific Rim countries include rapidly growing demand areas as well as significant reserves of gas. The region will continue to play a leading role in the evolution of a world trade in gas. Gas resources within the Asian Pacific region are adequate to serve the foreseeable demands, but historically the region has utilized liquefied natural gas (LNG) imports. Financial constraints upon the gas producing countries of the region and political instability in some of them will probably continue to require the importing of sustantial quantities of gas from the Middle East and possibly from Alaska and the former USSR as the resources indigenous to the region itself are developed more slowly than demand. The financial arrangements and contractual approaches that evolve to meet the needs of the Asia Pacific Rim will shape the future of world LNG markets. (Author)

  10. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

  11. Mobile Monitoring of Methane During and After the Aliso Canyon Natural Gas Leak

    Science.gov (United States)

    Polidori, A.; Pikelnaya, O.; Low, J.; Wimmer, R.; Zhou, Q.

    2016-12-01

    The Aliso Canyon gas leak was discovered inside the SoCalGas (SCG) facility on October 23, 2015. This incident represented the worst natural gas leak in the US history, and spurred a number of odor nuisance complaints from local residents. The community of Porter Ranch, located directly south of the SCG Aliso Canyon facility, was the most affected by the leak although complaints have been also reported in other neighboring communities of the San Fernando Valley. Therefore, monitoring of air quality was and remains crucial for measuring the impact of methane emissions from this leak and assessing the well-being of all residents. As the main local air quality agency for this area, South Coast Air Quality Management District (SCAQMD) organized a set of monitoring activities in response to the leak. Since December 21, 2015 SCAQMD has been conducting mobile survey measurements in and around Porter Ranch to characterize methane levels and concentration gradients within the community. For this purpose, a fast-response optical methane analyzer (LI-COR 7700) and a Global Positioning System (GPS) were mounted on top of a hybrid vehicle and driven around Porter Ranch and other surrounding areas. Following the permanent seal of the leaking well on February 18, 2016 mobile measurements have also been expanded to inside the Aliso Canyon SCG facility. During this presentation we will describe the experimental setup designed for mobile methane surveys and the monitoring strategy used for this study. We will discuss the main results of our mobile measurements including long-term methane trends since the end of the leak.

  12. Occurrence of gas hydrate in Oligocene Frio sand: Alaminos Canyon Block 818: Northern Gulf of Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Boswell, R.D.; Shelander, D.; Lee, M.; Latham, T.; Collett, T.; Guerin, G.; Moridis, G.; Reagan, M.; Goldberg, D.

    2009-07-15

    A unique set of high-quality downhole shallow subsurface well log data combined with industry standard 3D seismic data from the Alaminos Canyon area has enabled the first detailed description of a concentrated gas hydrate accumulation within sand in the Gulf of Mexico. The gas hydrate occurs within very fine grained, immature volcaniclastic sands of the Oligocene Frio sand. Analysis of well data acquired from the Alaminos Canyon Block 818 No.1 ('Tigershark') well shows a total gas hydrate occurrence 13 m thick, with inferred gas hydrate saturation as high as 80% of sediment pore space. Average porosity in the reservoir is estimated from log data at approximately 42%. Permeability in the absence of gas hydrates, as revealed from the analysis of core samples retrieved from the well, ranges from 600 to 1500 millidarcies. The 3-D seismic data reveals a strong reflector consistent with significant increase in acoustic velocities that correlates with the top of the gas-hydrate-bearing sand. This reflector extends across an area of approximately 0.8 km{sup 2} and delineates the minimal probable extent of the gas hydrate accumulation. The base of the inferred gas-hydrate zone also correlates well with a very strong seismic reflector that indicates transition into units of significantly reduced acoustic velocity. Seismic inversion analyses indicate uniformly high gas-hydrate saturations throughout the region where the Frio sand exists within the gas hydrate stability zone. Numerical modeling of the potential production of natural gas from the interpreted accumulation indicates serious challenges for depressurization-based production in settings with strong potential pressure support from extensive underlying aquifers.

  13. Mineral resources of the Desolation Canyon, Turtle Canyon, and Floy Canyon Wilderness Study Areas, Carbon Emery, and Grand counties, Utah

    International Nuclear Information System (INIS)

    Cashion, W.B.; Kilburn, J.E.; Barton, H.N.; Kelley, K.D.; Kulik, D.M.; McDonnell, J.R.

    1990-09-01

    This paper reports on the Desolation Canyon, Turtle Canyon, and Floy Canyon Wilderness Study Areas which include 242,000 acres, 33,690 acres, and 23,140 acres. Coal deposits underlie all three study areas. Coal zones in the Blackhawk and Nelsen formations have identified bituminous coal resources of 22 million short tons in the Desolation Canyon Study Area, 6.3 million short tons in the Turtle Canyon Study Area, and 45 million short tons in the Floy Canyon Study Area. In-place inferred oil shale resources are estimated to contain 60 million barrels in the northern part of the Desolation Canyon area. Minor occurrences of uranium have been found in the southeastern part of the Desolation Canyon area and in the western part of the Floy Canyon area. Mineral resource potential for the study areas is estimated to be for coal, high for all areas, for oil and gas, high for the northern tract of the Desolation Canyon area and moderate for all other tracts, for bituminous sandstone, high for the northern part of the Desolation Canyon area, and low for all other tracts, for oil shale, low in all areas, for uranium, moderate for the Floy Canyon area and the southeastern part of the Desolation Canyon area and low for the remainder of the areas, for metals other than uranium, bentonite, zeolites, and geothermal energy, low in all areas, and for coal-bed methane unknown in all three areas

  14. 75 FR 8152 - Pacific Gas and Electric Company; Diablo Canyon Power Plant Environmental Assessment and Finding...

    Science.gov (United States)

    2010-02-23

    ... exposures to plant workers and members of the public. Therefore, no changes or different types of... impacts to historical and cultural resources. There would be no impact to socioeconomic resources...

  15. Removal of tritium from gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1976-01-01

    Tritium contained in the coolant gas in the primary circuit of a gas cooled nuclear reactor together with further tritium adsorbed on the graphite used as a moderator for the reactor is removed by introducing hydrogen or a hydrogen-containing compound, for example methane or ammonia, into the coolant gas. The addition of the hydrogen or hydrogen-containing compound to the coolant gas causes the adsorbed tritium to be released into the coolant gas and the tritium is then removed from the coolant gas by passing the mixture of coolant gas and hydrogen or hydrogen-containing compound through a gas purification plant before recirculating the coolant gas through the reactor. 14 claims, 1 drawing figure

  16. Cruise report for a seismic investigation of gas hydrates in the Mississippi Canyon region, northern Gulf of Mexico; cruise M1-98-GM

    Science.gov (United States)

    Cooper, Alan K.; Hart, Patrick E.; Pecher, Ingo

    1998-01-01

    During June 1998, the U.S. Geological Survey (USGS) and the University of Mississippi Marine Minerals Technology Center (MMTC) conducted a 12-day cruise in the Mississippi Canyon region of the Gulf of Mexico (Fig. 1). The R/V Tommy Munro, owned by the Marine Research Institute of the University of Southern Mississippi, was chartered for the cruise. The general objective was to acquire very high resolution seismic-reflection data across of the upper and middle continental slope (200-1200-m water depths) to study the acoustic character, distribution and potential effects of gas hydrates within the shallow subsurface, extending from the sea floor down to the base of the gas-hydrate stability zone. The Gulf of Mexico is well known for hydrocarbon resources that include petroleum and related gases. Areas of the Gulf that lie in waters deeper than about 250 m potentially have conditions (e.g., pressure, temperature, near-surface gas content, etc.) that are right for the shallow-subsurface formation of the ice-like substance (gas and water) known as gas hydrate (Kvenvolden, 1993). Gas hydrates have previously been sampled in sea-floor cores and observed as massive mounds in several parts of the northern Gulf, including the Mississippi Canyon region (e.g., Anderson et al., 1992). Extensive seismic data have been recorded in the Gulf, in support of commercial drilling efforts, but few very high resolution data exist in the public domain to aid in gas-hydrate studies. Studies of long-term interest include those on the resource potential of gas hydrates, the geologic hazards associated with dissociation and formation of hydrates, and the impact, if any, of gas-hydrate dissociation on atmospheric warming (i.e., via release of methane, a "greenhouse" gas). Several very high resolution seismic systems (surface-towed, deep-towed, and sea-floor) were used during the cruise to test the feasibility of using such data for detailed structural (geometric) and stratigraphic (physical

  17. Pacific Northern Gas Ltd. annual report 2002

    International Nuclear Information System (INIS)

    2003-01-01

    Pacific Northern Gas Ltd. operates in west-central and northeast British Columbia. The company delivers natural gas to customers through a transmission pipeline connected to Duke Energy system near Summit Lake, British Columbia. This report states that in 2002 financial results were disappointing. The company's net income in 2002 was lower than it was in 2001 ($4.6 million versus $5.7 million). In December 2002, Pacific Northern Gas Ltd. completed $15 million in financing. Additions to property, plant and equipment reached a total of $6 million in 2002. A new, seven-year contract with Methanex Corporation was successfully negotiated. Pacific Northern Gas Ltd. filed revenue requirements applications with the British Columbia Utilities Commission, seeking the Commission's approval of rates for 2003 and requesting approval of a new deferral account in all divisions. A settlement of the western system 2003 revenue requirements application was negotiated with its customers. The annual report presented a highlight of all activities, including corporate governance and management discussions and analysis. Consolidated financial statements were also provided. tabs

  18. 236-Z canyon utilization study

    International Nuclear Information System (INIS)

    Dixon, D.R.

    1977-01-01

    The 236-Z canyon contains equipment for repurification of plutonium and recovery of plutonium from scrap material. To meet production requirements of Fast Flux Test Facility/Clinch River Breeder Reactor oxide with the existing plant, several new pieces of equipment will be needed in the future. More storage space and a better accountability system are needed to support this increased production. The available canyon space needs to be utilized to its fullest in order to accommodate the new equipment. The purpose of this document is to identify the new pieces of equipment, show how they fit into the flowsheet, and locate them in the canyon

  19. Level 1 probabilistic risk assessment of low power and shutdown operations at a PWR: Phase 2 results

    International Nuclear Information System (INIS)

    Chu, T.L.; Bozoki, G.; Kohut, P.; Musicki, Z.; Wong, S.M.; Yang, J.; Hsu, C.J.; Diamond, D.J.; Su, R.F.; Holmes, B.; Siu, N.; Bley, D.; Lin, J.

    1992-01-01

    As a result of the Chernobyl accident and other precursor events (e.g., Diablo Canyon), the US Nuclear Regulatory Commission's (NRC's) Office of Nuclear Regulatory Research (RES) initiated an extensive project during 1989 to carefully examine the potential risks during Low Power and Shutdown (LP ampersand S) operations. Shortly after the program began, an event occurred at the Vogtle plant during shutdown, which further intensified the effort of the LP ampersand S program. In the LP ampersand S program, one pressurized water reactor (PWR), Surry, and one boiling water reactor (BWR), Grand Gulf, were selected, mainly because they were previously analyzed in the NUREG-1150 Study. The Level-1 Program is being performed in two phases. Phase 1 was dedicated to performing a coarse screening level-1 analysis including internal fire and flood. A draft report was completed in November, 1991. In the phase 2 study, mid-loop operations at the Surry plant were analyzed in detail. The objective of this paper is to present the approach of the phase 2 study and the preliminary results and insights

  20. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    Bastien, D.

    1996-01-01

    The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

  1. The Asia Pacific natural gas market: Large enough for all?

    International Nuclear Information System (INIS)

    Aguilera, Roberto F.; Inchauspe, Julian; Ripple, Ronald D.

    2014-01-01

    Among natural gas producing nations, there has been some concern about how the Asia Pacific will meet future demand for energy. We argue that natural gas, both regional and global, will play a vital role. Estimates of potential gas consumption in the region are analyzed and used to develop consensus projections to 2030. These consumption profiles are compared with gas supply estimates including indigenous, pipeline and LNG for the Asia Pacific market. From this analytical framework, we find that demand will be sufficiently large to accommodate supplies from diverse sources including North America, the Middle East, Central Asia, Russia, and the Asia Pacific itself. An important policy implication is that gas producing and consuming nations should benefit from promoting gas trade and not be concerned about a situation of potential lack of demand coupled with oversupply. - Highlights: • Estimates of gas consumption in the Asia Pacific (AP) in 2030 are presented. • Compared with supply estimates for AP including indigenous, pipeline, and LNG. • Find that demand in AP large enough to accommodate supply from all regions. • Nations should promote gas trade policy and not be overly concerned about oversupply

  2. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  3. Dissolution of Material and Test reactor Fuel in an H-Canyon Dissolver

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-26

    In an amended record of decision for the management of spent nuclear fuel (SNF) at the Savannah River Site, the US Department of Energy has authorized the dissolution and recovery of U from 1000 bundles of Al-clad SNF. The SNF is fuel from domestic and foreign research reactors and is typically referred to as Material Test Reactor (MTR) fuel. Bundles of MTR fuel containing assemblies fabricated from U-Al alloys (or other U compounds) are currently dissolved using a Hg-catalyzed HNO3 flowsheet. Since the development of the existing flowsheet, improved experimental methods have been developed to more accurately characterize the offgas composition and generation rate during laboratory dissolutions. Recently, these new techniques were successfully used to develop a flowsheet for the dissolution of High Flux Isotope Reactor (HFIR) fuel. Using the data from the HFIR dissolution flowsheet development and necessary laboratory experiments, the Savannah River National Laboratory (SRNL) was requested to define flowsheet conditions for the dissolution of MTR fuels. With improved offgas characterization techniques, SRNL will be able define the number of bundles of fuel which can be charged to an H-Canyon dissolver with much less conservatism.

  4. Monitoring ground-surface heating during expansion of the Casa Diablo production well field at Mammoth Lakes, California

    Science.gov (United States)

    Bergfeld, D.; Vaughan, R. Greg; Evans, William C.; Olsen, Eric

    2015-01-01

    The Long Valley hydrothermal system supports geothermal power production from 3 binary plants (Casa Diablo) near the town of Mammoth Lakes, California. Development and growth of thermal ground at sites west of Casa Diablo have created concerns over planned expansion of a new well field and the associated increases in geothermal fluid production. To ensure that all areas of ground heating are identified prior to new geothermal development, we obtained high-resolution aerial thermal infrared imagery across the region. The imagery covers the existing and proposed well fields and part of the town of Mammoth Lakes. Imagery results from a predawn flight on Oct. 9, 2014 readily identified the Shady Rest thermal area (SRST), one of two large areas of ground heating west of Casa Diablo, as well as other known thermal areas smaller in size. Maximum surface temperatures at 3 thermal areas were 26–28 °C. Numerous small areas with ground temperatures >16 °C were also identified and slated for field investigations in summer 2015. Some thermal anomalies in the town of Mammoth Lakes clearly reflect human activity.Previously established projects to monitor impacts from geothermal power production include yearly surveys of soil temperatures and diffuse CO2 emissions at SRST, and less regular surveys to collect samples from fumaroles and gas vents across the region. Soil temperatures at 20 cm depth at SRST are well correlated with diffuse CO2 flux, and both parameters show little variation during the 2011–14 field surveys. Maximum temperatures were between 55–67 °C and associated CO2 discharge was around 12–18 tonnes per day. The carbon isotope composition of CO2 is fairly uniform across the area ranging between –3.7 to –4.4 ‰. The gas composition of the Shady Rest fumarole however has varied with time, and H2S concentrations in the gas have been increasing since 2009.

  5. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  6. Compilation of PRF Canyon Floor Pan Sample Analysis Results

    Energy Technology Data Exchange (ETDEWEB)

    Pool, Karl N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Minette, Michael J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wahl, Jon H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Greenwood, Lawrence R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Coffey, Deborah S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); McNamara, Bruce K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Bryan, Samuel A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Scheele, Randall D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Delegard, Calvin H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sinkov, Sergey I. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Soderquist, Chuck Z. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Fiskum, Sandra K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Brown, Garrett N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Clark, Richard A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-06-30

    On September 28, 2015, debris collected from the PRF (236-Z) canyon floor, Pan J, was observed to exhibit chemical reaction. The material had been transferred from the floor pan to a collection tray inside the canyon the previous Friday. Work in the canyon was stopped to allow Industrial Hygiene to perform monitoring of the material reaction. Canyon floor debris that had been sealed out was sequestered at the facility, a recovery plan was developed, and drum inspections were initiated to verify no additional reactions had occurred. On October 13, in-process drums containing other Pan J material were inspected and showed some indication of chemical reaction, limited to discoloration and degradation of inner plastic bags. All Pan J material was sealed back into the canyon and returned to collection trays. Based on the high airborne levels in the canyon during physical debris removal, ETGS (Encapsulation Technology Glycerin Solution) was used as a fogging/lock-down agent. On October 15, subject matter experts confirmed a reaction had occurred between nitrates (both Plutonium Nitrate and Aluminum Nitrate Nonahydrate (ANN) are present) in the Pan J material and the ETGS fixative used to lower airborne radioactivity levels during debris removal. Management stopped the use of fogging/lock-down agents containing glycerin on bulk materials, declared a Management Concern, and initiated the Potential Inadequacy in the Safety Analysis determination process. Additional drum inspections and laboratory analysis of both reacted and unreacted material are planned. This report compiles the results of many different sample analyses conducted by the Pacific Northwest National Laboratory on samples collected from the Plutonium Reclamation Facility (PRF) floor pans by the CH2MHill’s Plateau Remediation Company (CHPRC). Revision 1 added Appendix G that reports the results of the Gas Generation Rate and methodology. The scope of analyses requested by CHPRC includes the determination of

  7. TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

    International Nuclear Information System (INIS)

    Grover, S.B.

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  8. Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    S. Blaine Grover

    2004-01-01

    The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

  9. Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

    Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

  10. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High Temperature Reactor (VHTR), as well as actinide burning concepts [ ]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic ‘honeycomb’ structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  11. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    Maturana, Roberto Hernan

    2008-01-01

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design. [es

  12. Advanced gas-cooled reactors (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

    1981-01-15

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

  13. 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

  14. Research reactor collaboration in the Asia-Pacific region

    International Nuclear Information System (INIS)

    Jun, Byung Jin

    2006-01-01

    The number of research reactors over the world has been decreasing since its peak in the middle of the 1970s, and it is predicted to decrease more rapidly than before in the future. International collaboration on research reactors is an effective way for their continued safe service to human welfare in various technical areas. The number of new research reactors under construction or planned for in the Asia-Pacific region is the greatest in the world. Among the regional collaboration activities on research reactors, safety has been the most important subject followed by neutron activation analysis, radioisotope production and neutron beam applications. It is understood that more regional collaboration on basic technologies important for the safety, management and utilization of the research reactors is demanding. The new project proposal of the Forum for Nuclear Cooperation in Asia on 'Research Reactor Technology for Effective Utilization' is understood to meet the demands. Meanwhile, there is a consensus on the need for research reactor resource sharing in the region. As a result of the review on the international collaboration activities in the region, the author suggests a linkage between the above new project and IAEA/RCA project considering a possible sharing of research reactor resources in the region. (author)

  15. Integrated approach to natural gas utilization in the Asia Pacific region

    International Nuclear Information System (INIS)

    Hovdestad, W.R.; Egbogah, E.O.

    1995-01-01

    The rapidly expanding economies in the Pacific Rim have placed increasing demands upon indigenous natural gas supplies in South East Asia and Australia. Competing demands include exports of liquefied natural gas (LNG), domestic consumption, and potential use for enhanced oil recovery (EOR) to extend the useful life of maturing oil fields. An additional competing demand for gas exports may emerge as the interstate pipeline grid is expanded. An integrated approach incorporating the evolving nature of gas demands and discrete physical supplies would provide a means to mitigate against potential mismatching of supply and demand. The consideration of the evolving nature of gas demands could promote economically beneficial changes to gas field development. The development of high carbon dioxide (CO 2 ) content gas fields has been slowed by the lack of a market for CO 2 . Utilization of by-product CO 2 for EOR could improve development economics, thus facilitating earlier development of gas supplies to satisfy gas demands including domestic use and LNG exports. End users would also benefit from the assurance that gas supplies would become available as needed. The maturity and increasingly complex natural gas industry in the Asia Pacific Region has led to a qualitative change. The model of single projects to satisfy single markets is no longer valid. The current environment is more dynamic, creating the need to anticipate changes to market demands and to find value-added markets for by-products. The integrated approach to gas utilization discussed in this paper presents a new model more appropriate to the gas industry existing today in the Asia Pacific Region. This approach is particularly significant to widely discussed proposals for an Asia Pacific energy grid extending to Australia

  16. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki

    1982-07-01

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  17. Scram device for gas-cooled reactor

    International Nuclear Information System (INIS)

    Murakami, Atsushi; Takahashi, Suehiro.

    1989-01-01

    A scram device for gas-cooled reactors has a hopper disposed below a stand pipe standing upright passing through a reactor container and electromagnets disposed therein. It further comprises neutron absorbing steel balls maintained between the electromagnets and the hopper upon energization of the electromagnets. Upon emergency reactor shutdown, energization for the electromagnets is interrupted to drop the neutron absorption stainless steel balls into the reactor core. It is an object of the present invention to keep the mechanical strength of the electromagnets in a high temperature gas atmosphere and not to reduce the insulation performance. That is, coils for the electromagnets are constituted with a small oxide-insulated metal sheath cable (MI cable). As the feature of the MI cable, it can maintain the mechanical strength even when exposed to high temperature gas coolant and the insulation performance thereof does not reduce by virture of its gas sealing property. Accordingly, a scram device of stable reliability can be obtained. (K.M.)

  18. Predicted radionuclide release from reactor-related unenclosed solid objects dumped in the Sea of Japan and the Pacific Ocean, east coast of Kamchatka

    International Nuclear Information System (INIS)

    Mount, M.E.; Lynn, N.M.; Warden, J.M.

    1996-06-01

    Between 1978 and 1991 reactor-related solid radioactive waste was dumped by the former Soviet Union as unenclosed objects in the Pacific Ocean, east coast of Kamchatka, and the Sea of Japan. This paper presented estimates for the current (1994) inventory of activation and corrosion products contained in the reactor-related unenclosed solid objects. In addition, simple models derived for prediction of radionuclide release from marine reactors dumped in the Kara Sea are applied to certain of the dumped objects to provide estimates of radionuclide release to the Pacific Ocean, east coast of Kamchatka, and Sea of Japan environments. For the Pacific Ocean, east coast of Kamchatka, total release rates start below 0.01 GBq yr -1 and over 1,000 years, fall to 100 Bq yr -1 . In the Sea of Japan, the total release rate starts just above 1 GBq yr - 1 , dropping off to a level less than 0.1 GBq yr -1 , extending past the year 4,000

  19. 76 FR 24465 - Pacific Gas and Electric Company

    Science.gov (United States)

    2011-05-02

    ... Electric Company Notice of Application Tendered for Filing With the Commission and Establishing Procedural... Gas and Electric Company. e. Name of Project: Drum-Spaulding Project. f. Location: The west slope of..., Pacific Gas and Electric Company, P.O. Box 770000, San Francisco, CA 94177-0001, (415) 973-4481, or e-mail...

  20. Estimation of immersion dose to the Tin mining in the seashore of Bangka island from NPP operation

    International Nuclear Information System (INIS)

    Nurokhim; Erwansyah Lubis

    2015-01-01

    The estimation of immersion dose to the Tin (Sn) mining in the seashore of Bangka island using concentrations factor methods was carried out. In the estimation, the source-term of effluent released to Pacific ocean from Diablo Canyon and San Onofre Nuclear Power Plant (NPP) operations was used. The results indicated that the Sn mining with immersion in the sea water of 2, 4 and 6 hours per day will receives maximum effective dose of 4.45 × 10 -3 %; 8.90 ×10 -3 % and 1.34 × 10 -2 % from dose constraint of 0.3 mSv per years. The probability of cancer happen for individual are 2.67 × 10 -8 ; 5.34 × 10 -8 and 8.01 × 10 -8 respectively if the working hours are 2, 4 dan 6 hours per days as long as 40 years (reactor lifetime) . These data give the information if there are 50 millions of Sn mining, the potential of miner will receive a fatal cancer is around 4 persons. According to the population of Sn mining is very small, so the probabilities of fatal cancer is unsignificant. (author)

  1. Natural gas supply and demand projections for the Asia-Pacific region

    International Nuclear Information System (INIS)

    Khin, J.A.

    1992-01-01

    The phenomenon of rapid economic growth in the Asia Pacific has inevitably led the countries of this region to expand and diversify their energy sources in order to satisfy their burgeoning energy demands. Natural gas has become an increasingly marketable energy source in this region benefitting from vast reserves and its advantages as an environmentally clean fuel. As a result of the impact of the two oil shocks of the 1970's on the Asia Pacific economy, the governments in the region set about the development of energy strategies which would make their national economies more resilient to the instabilities of world energy price and supply. The Japanese Gas Industry has estimated that the overall rise in demand for energy in Asia, set at an average rate of 3.7% per annum, will see a corresponding growth in demand for natural gas at 5%. Experts from a number of major oil companies, such as Exxon, expect an annual growth in the Asian natural gas market of 6.0 to 6.3%. These figures are over shadowed by the worldwide demand for natural gas which is expected to gain an 8% increase within the next two decades. Approximately 8.75% of the world's proven natural gas reserves are held in Asia-Pacific region (Table I). Most of the region's natural gas production will provide over 100 years of supply. A review of natural gas supply/demand in the Asia-Pacific region is presented in sub-regions, namely ASIAN, Northeast Asia, the Indian Subcontinent and Oceania

  2. Plant experience with temporary reverse osmosis makeup water systems

    International Nuclear Information System (INIS)

    Polidoroff, C.

    1986-01-01

    Pacific Gas and Electric (PG and E) Company's Diablo Canyon Power Plant (DCPP), which is located on California's central coast, has access to three sources of raw water: creek water, well water, and seawater. Creek and well water are DCPP's primary sources of raw water; however, because their supply is limited, these sources are supplemented with seawater. The purpose of this paper is to discuss the temporary, rental, reverse osmosis systems used by PG and E to process DCPP's raw water into water suitable for plant makeup. This paper addresses the following issues: the selection of reverse osmosis over alternative water processing technologies; the decision to use vendor-operated temporary, rental, reverse osmosis equipment versus permanent PG and E-owned and -operated equipment; the performance of DCPP's rental reverse osmosis systems; and, the lessons learned from DCPP's reverse osmosis system rental experience that might be useful to other plants considering renting similar equipment

  3. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    1985-12-01

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

  4. Neutronics of a mixed-flow gas-core reactor

    International Nuclear Information System (INIS)

    Soran, P.D.; Hansen, G.E.

    1977-11-01

    The study was made to investigate the neutronic feasibility of a mixed-flow gas-core reactor. Three reactor concepts were studied: four- and seven-cell radial reactors and a seven-cell scallop reactor. The reactors were fueled with UF 6 (either U-233 or U-235) and various parameters were varied. A four-cell reactor is not practical nor is the U-235 fueled seven-cell radial reactor; however, the 7-cell U-233 radial and scallop reactors can satisfy all design criteria. The mixed flow gas core reactor is a very attractive reactor concept and warrants further investigation

  5. Fire damp gas in a heavy water reactor; Praskavi gas u teskovodnom reaktoru

    Energy Technology Data Exchange (ETDEWEB)

    Nikolic, V D [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Yugoslavia)

    1963-07-01

    This document describes the process of fire damp gas creation in the reactor core and dependence of the gas percentage on the temperature, i.e. reactor power. It contains a detailed plan for measuring the the percent of fire damp gas at the RA reactor: before start-up, after longer shut-down periods, immediately after safety shutdown, periodically during operation campaign.

  6. Using size distribution analysis to forecast natural gas resources in Asia Pacific

    International Nuclear Information System (INIS)

    Aguilera, Roberto F.; Ripple, Ronald D.

    2011-01-01

    Highlights: → We estimate the total endowment of conventional natural gas in Asia Pacific. → Includes volumes in previously unassessed provinces. → Endowment distributed across countries to show where volumes are most likely to be found. → A breakdown between offshore versus onshore resources is also estimated. → We find there is a significant natural gas endowment in the region. -- Abstract: Increasing energy consumption in Asia Pacific will largely be met by fossil fuels. Natural gas production in the region presently ranks behind that of oil and coal. However, the abundance of gas could lead to a significant gas market share increase in the energy mix. The purpose of this paper is to estimate the total endowment of conventional gas in Asia Pacific. This is carried out with a Variable Shape Distribution (VSD) model that forecasts volumes in provinces that have not been previously evaluated. The endowment is then distributed across countries to show where volumes are most likely to be found. A breakdown between offshore versus onshore resources is also estimated. The results of the analysis show there is a significant gas endowment. The estimated distribution across countries and onshore/offshore areas provides insight into the relative economics of gas production, as well as a basis for potential investment decisions. With appropriate energy policies, it may be possible to tap the vast gas potential in Asia Pacific. Considering gas may be the most abundant, inexpensive, and clean fossil fuel, the outcome would be increased energy security and a low carbon economy.

  7. Therapeutic Potential of a Novel Smac/DIABLO in Breast Cancer

    National Research Council Canada - National Science Library

    Srivastava, Rakesh K

    2005-01-01

    ... potential of tamoxifen, doxorubicin and paclitaxel. Overexpression of Smac/DIABLO gene or Smac peptide enhances the apoptosis-inducing potential of tamoxifen, doxorubicin and paclitaxel, and sentitizes TRAiL-resistant breast cancer cell lines...

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

    International Nuclear Information System (INIS)

    1978-09-01

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

  9. Experimental simulation of air quality in street canyon under changes of building orientation and aspect ratio.

    Science.gov (United States)

    Yassin, Mohamed F; Ohba, Masaake

    2012-09-01

    To assist validation of numerical simulations of urban pollution, air quality in a street canyon was investigated using a wind tunnel as a research tool under neutral atmospheric conditions. We used tracer gas techniques from a line source without buoyancy. Ethylene (C(2)H(4)) was used as the tracer gas. The street canyon model was formed of six parallel building rows of the same length. The flow and dispersion field was analyzed and measured using a hot-wire anemometer with split fiber probe and fast flame ionization detector. The diffusion flow field in the boundary layer within the street canyon was examined at different locations, with varying building orientations (θ=90°, 112.5°, 135° and 157.5°) and street canyon aspect ratios (W/H=1/2, 3/4 and 1) downwind of the leeward side of the street canyon model. Results show that velocity increases with aspect ratio, and with θ>90°. Pollutant concentration increases as aspect ratio decreases. This concentration decreases exponentially in the vertical direction, and decreases as θ increases from 90°. Measured pollutant concentration distributions indicate that variability of building orientation and aspect ratio in the street canyon are important for estimating air quality in the canyon. The data presented here can be used as a comprehensive database for validation of numerical models.

  10. Ethanolamine properties and use for feedwater pH control: A pressurized water reactor case study

    International Nuclear Information System (INIS)

    Keeling, D.L.; Polidoroff, C.T.; Cortese, S.; Cushner, M.C.

    1995-01-01

    Ethanolamine (ETA) as a feedwater pH control additive has been recently used to minimize corrosion of secondary water components in the nuclear power industry pressurized water reactors (PWRs). The use of ETA is compared with ammonia. Relative volatility effects on various parts of the system are analyzed and chemistry changes are presented. Materials of construction and the use of existing plant equipment for ETA service are discussed. Properties of ETA as well as safety, storage and handling issues are compared with ammonia. Health d aquatic toxicity are reviewed. warnings, safety, handling guidelines, biodegradability an Diablo Canyon Power Plant used ammonia for pH control from 1985 until a change over to ETA in 1993/1994. Full flow condensate polishers that are required to protect the plant from saltwater cooling incursions limit the amount of pH additive. Iron levels in the secondary water systems are compared before and after changing to ETA and replacement of corrosion-susceptible piping. Iron reduction benefits are assessed along with other effects on the feedwater nozzles, low pressure turbine, polisher resin capacity and polisher regeneration system

  11. Organic geochemical investigation and coal-bed methane characteristics of the Guasare coals (Paso Diablo mine, western Venezuela)

    Science.gov (United States)

    Quintero, K.; Martinez, M.; Hackley, P.; Marquez, G.; Garban, G.; Esteves, I.; Escobar, M.

    2011-01-01

    The aim of this work was to carry out a geochemical study of channel samples collected from six coal beds in the Marcelina Formation (Zulia State, western Venezuela) and to determine experimentally the gas content of the coals from the Paso Diablo mine. Organic geochemical analyses by gas chromatography-mass spectrometry and isotopic analyses on-line in coalbed gas samples were performed. The results suggest that the Guasare coals were deposited in a continental environment under highly dysoxic and low salinity conditions. The non-detection of 18??(H)-oleanane does not preclude that the organic facies that gave rise to the coals were dominated by angiosperms. In addition, the presence of the sesquiterpenoid cadalene may indicate the subordinate contribution of gymnosperms (conifers) in the Paleocene Guasare mire. The average coalbed gas content obtained was 0.6 cm3/g. ??13C and D values indicate that thermogenic gas is prevalent in the studied coals. Copyright ?? Taylor & Francis Group, LLC.

  12. IAEA high temperature gas-cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2000-01-01

    The IAEA activities on high temperature gas-cooled reactors are conducted with the review and support of the Member states, primarily through the International Working Group on Gas-Cooled Reactors (IWG-GCR). This paper summarises the results of the IAEA gas-cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (authors)

  13. IAEA high temperature gas cooled reactor activities

    International Nuclear Information System (INIS)

    Kendall, J.M.

    2001-01-01

    IAEA activities on high temperature gas cooled reactors are conducted with the review and support of Member States, primarily through the International Working Group on Gas Cooled Reactors (IWGGCR). This paper summarises the results of the IAEA gas cooled reactor project activities in recent years along with ongoing current activities through a review of Co-ordinated Research Projects (CRPs), meetings and other international efforts. A series of three recently completed CRPs have addressed the key areas of reactor physics for LEU fuel, retention of fission products, and removal of post shutdown decay heat through passive heat transport mechanisms. These activities along with other completed and ongoing supporting CRPs and meetings are summarised with reference to detailed documentation of the results. (author)

  14. Nuclear interaction of Smac/DIABLO with Survivin at G2/M arrest prompts docetaxel-induced apoptosis in DU145 prostate cancer cells

    International Nuclear Information System (INIS)

    Kim, Ji Young; Chung, Jin-Yong; Lee, Seung Gee; Kim, Yoon-Jae; Park, Ji-Eun; Yoo, Ki Soo; Yoo, Young Hyun; Park, Young Chul; Kim, Byeong Gee; Kim, Jong-Min

    2006-01-01

    Smac/DIABLO is released by mitochondria in response to apoptotic stimuli and is thought to antagonize the function of inhibitors of apoptosis proteins. Recently, it has been shown that, like XIAP, Survivin can potentially interact with Smac/DIABLO. However, the precise mechanisms and cellular location of their action have not been determined. We report for the first time that Smac/DIABLO translocates to the nucleus and is colocalized with Survivin at mitotic spindles during apoptosis resulting from G2/M arrest due to docetaxel treatment of DU145 prostate cancer cells. Our data demonstrate that the nuclear interaction of Smac/DIABLO with Survivin is an important step for suppressing the anti-apoptotic function of Survivin in Doc-induced apoptosis. This suggests that the balance between cellular Smac/DIABLO and Survivin levels could be critical for cellular destiny in taxane-treated cancer cells

  15. In-situ calibration of nuclear-plant platinum resistance thermometers using Johnson noise methods. Final report

    International Nuclear Information System (INIS)

    Blalock, T.V.; Shepard, R.L.

    1983-06-01

    Methods for in situ calibration of reactor plant platinum resistance thermometers using Johnson noise measurements were tested in the laboratory and in two operating reactors: Diablo Canyon and Sequoyah. The Johnson noise methods provide an absolute measurement of the thermometer temperature and can be compared with the dc calibration of the thermometers to confirm their calibration without removing the thermometers from the plant coolant loops. Inaccuracies of less than 0.1% were obtained with these methods for ideal situations where the noise measuring equipment could be connected directly to the installed thermometer terminals. For measurements made at the ends of long extension cables, inaccuracies were 0.5 to 1.0%, at best. Extension cables must be optimized and well characterized electrically to achieve such accuracies. Other factors that affect the accuracy of these methods were evaluated

  16. Eighth meeting of the International Working Group on Gas-Cooled Reactors Vienna, 30 January - 1 February 1989. Summary report. Part 2

    International Nuclear Information System (INIS)

    1989-12-01

    The Eighth Meeting of the IAEA International Working Group on Gas-Cooled Reactors was held in Vienna, Austria, from 30 January - 1 February, 1989. The Summary Report (Part II) contains the papers which review the national programmes in the field of Gas-Cooled Reactors and other presentations at the Meeting. Refs, figs and tabs

  17. Gas cooled reactors

    International Nuclear Information System (INIS)

    Kojima, Masayuki.

    1985-01-01

    Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

  18. Gas fluidized bed reactor

    International Nuclear Information System (INIS)

    Bernardelli, H. da C.

    1976-03-01

    The equations of motion for both gas and particles in a gas fluidised system are stablished through general assumptions which are generally accepted on physical grounds. The resulting model is used to study the velocity fields of each phase in the case of an isolated bubble rising close to the flat distributor plate. A well posed problem results for the solution of Laplace's equation of the potential flow of the particles when consideration is given to the presence of the distributor as a boundary condition. The corresponding stream functions are also obtained which enable the drawing of the motion patterns using numerical techniques. The following two dimensional cases are analysed: S/b=1; S/b=1,5; S/b=2,5; S/b=5 and the limiting case S/b→αinfinite. The results for the interphase exchange between bubbles and particulate phases are applied to a gas fluidised bed reactor and its effect on the chemical conversion is studied for the simplest cases of piston flow and perfect mixing in the particulate phase [pt

  19. Technical specifications: Diablo Canyon Nuclear Power Plant, Units 1 and 2 (Docket Nos. 50-275 and 50-323)

    International Nuclear Information System (INIS)

    1985-08-01

    Safety limits and limiting safety system settings are presented for the reactivity control systems; power distribution limits; instrumentation; reactor coolant system; emergency core cooling system; containment systems; plant systems; electrical power systems; refueling operations; special test exceptions; radioactive effluents; and radiological environmental monitoring

  20. 76 FR 61687 - Pacific Gas and Electric Company

    Science.gov (United States)

    2011-10-05

    ... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Project No. 2479-011-CA] Pacific Gas and Electric Company Notice of Availability of Environmental Assessment In accordance with the National Environmental Policy Act of 1969 and the Federal Energy Regulatory Commission's (Commission...

  1. Isotropic, anisotropic, and borehole washout analyses in Gulf of Mexico Gas Hydrate Joint Industry Project Leg II, Alaminos Canyon well 21-A

    Science.gov (United States)

    Lee, Myung W.

    2012-01-01

    Through the use of three-dimensional seismic amplitude mapping, several gas hydrate prospects were identified in the Alaminos Canyon area of the Gulf of Mexico. Two of the prospects were drilled as part of the Gulf of Mexico Gas Hydrate Joint Industry Program Leg II in May 2009, and a suite of logging-while-drilling logs was acquired at each well site. Logging-while-drilling logs at the Alaminos Canyon 21–A site indicate that resistivities of approximately 2 ohm-meter and P-wave velocities of approximately 1.9 kilometers per second were measured in a possible gas-hydrate-bearing target sand interval between 540 and 632 feet below the sea floor. These values are slightly elevated relative to those measured in the hydrate-free sediment surrounding the sands. The initial well log analysis is inconclusive in determining the presence of gas hydrate in the logged sand interval, mainly because large washouts in the target interval degraded well log measurements. To assess gas-hydrate saturations, a method of compensating for the effect of washouts on the resistivity and acoustic velocities is required. To meet this need, a method is presented that models the washed-out portion of the borehole as a vertical layer filled with seawater (drilling fluid). Owing to the anisotropic nature of this geometry, the apparent anisotropic resistivities and velocities caused by the vertical layer are used to correct measured log values. By incorporating the conventional marine seismic data into the well log analysis of the washout-corrected well logs, the gas-hydrate saturation at well site AC21–A was estimated to be in the range of 13 percent. Because gas hydrates in the vertical fractures were observed, anisotropic rock physics models were also applied to estimate gas-hydrate saturations.

  2. Eighth meeting of the International Working Group on Gas-Cooled Reactors, Vienna, 30 January - 1 February 1989. Summary report. Part 1

    International Nuclear Information System (INIS)

    1989-12-01

    The Eighth Meeting of the IAEA International Working Group on Gas-Cooled Reactors was held in Vienna, Austria, from 30 January - 1 February, 1989. The Summary Report (Part I) contains the Minutes of the Meeting

  3. Large eddy simulation of pollutant gas dispersion with buoyancy ejected from building into an urban street canyon.

    Science.gov (United States)

    Hu, L H; Xu, Y; Zhu, W; Wu, L; Tang, F; Lu, K H

    2011-09-15

    The dispersion of buoyancy driven smoke soot and carbon monoxide (CO) gas, which was ejected out from side building into an urban street canyon with aspect ratio of 1 was investigated by large eddy simulation (LES) under a perpendicular wind flow. Strong buoyancy effect, which has not been revealed before, on such pollution dispersion in the street canyon was studied. The buoyancy release rate was 5 MW. The wind speed concerned ranged from 1 to 7.5m/s. The characteristics of flow pattern, distribution of smoke soot and temperature, CO concentration were revealed by the LES simulation. Dimensionless Froude number (Fr) was firstly introduced here to characterize the pollutant dispersion with buoyancy effect counteracting the wind. It was found that the flow pattern can be well categorized into three regimes. A regular characteristic large vortex was shown for the CO concentration contour when the wind velocity was higher than the critical re-entrainment value. A new formula was theoretically developed to show quantitatively that the critical re-entrainment wind velocities, u(c), for buoyancy source at different floors, were proportional to -1/3 power of the characteristic height. LES simulation results agreed well with theoretical analysis. The critical Froude number was found to be constant of 0.7. Copyright © 2010 Elsevier B.V. All rights reserved.

  4. Medium temperature carbon dioxide gas turbine reactor

    International Nuclear Information System (INIS)

    Kato, Yasuyoshi; Nitawaki, Takeshi; Muto, Yasushi

    2004-01-01

    A carbon dioxide (CO 2 ) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 deg. C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 deg. C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO 2 ; and consideration of variation in CO 2 specific heat at constant pressure, C p , with pressure and temperature into cycle configuration. Lowering temperature to 650 deg. C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 deg. C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO 2 have been proven during extensive operation in AGRs. In the previous study, the CO 2 cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO 2 gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors

  5. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Taryo, Taswanda

    2002-01-01

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  6. Organic molecules in the atmosphere of Jupiter. Final report

    International Nuclear Information System (INIS)

    Ponnamperuma, C.A.

    1978-01-01

    Organic synthesis in the primitive solar system was simulated by Fischer Tropsch type experiments. Particular attention was given to the formation of lower molecular weight hydrocarbons. In a gas flow experiment, a gas mixture of H 2 and CO was introduced into a heated reaction tube at a constant flow rate and passed through a catalyst (powdered Canyon Diablo). The products that emerged were directly analyzed by gas chromatography. The results of 21 runs under various gas mixing rations, reaction temperatures, and gas-catalyst contact times showed the predominance of the saturated hydrocarbon formation at C 4 and C 5 over the unsaturated ones. Saturate/unsaturate ratios were mostly less than 0.4 and none showed over 0.7

  7. Wind tunnel simulation of air pollution dispersion in a street canyon.

    Science.gov (United States)

    Civis, Svatopluk; Strizík, Michal; Janour, Zbynek; Holpuch, Jan; Zelinger, Zdenek

    2002-01-01

    Physical simulation was used to study pollution dispersion in a street canyon. The street canyon model was designed to study the effect of measuring flow and concentration fields. A method of C02-laser photoacoustic spectrometry was applied for detection of trace concentration of gas pollution. The advantage of this method is its high sensitivity and broad dynamic range, permitting monitoring of concentrations from trace to saturation values. Application of this method enabled us to propose a simple model based on line permeation pollutant source, developed on the principle of concentration standards, to ensure high precision and homogeneity of the concentration flow. Spatial measurement of the concentration distribution inside the street canyon was performed on the model with reference velocity of 1.5 m/s.

  8. Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

    International Nuclear Information System (INIS)

    Beech, D.J.; May, R.

    2000-01-01

    In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

  9. Seasonal to Mesoscale Variability of Water Masses in Barrow Canyon,Chukchi Sea

    Science.gov (United States)

    Nobre, C.; Pickart, R. S.; Moore, K.; Ashjian, C. J.; Arrigo, K. R.; Grebmeier, J. M.; Vagle, S.; Itoh, M.; Berchok, C.; Stabeno, P. J.; Kikuchi, T.; Cooper, L. W.; Hartwell, I.; He, J.

    2016-02-01

    Barrow Canyon is one of the primary conduits by which Pacific-origin water exits the Chukchi Sea into the Canada Basin. As such, it is an ideal location to monitor the different water masses through the year. At the same time, the canyon is an energetic environment where mixing and entrainment can occur, modifying the pacific-origin waters. As part of the Distributed Biological Observatory (DBO) program, a transect across the canyon was occupied 24 times between 2010-2013 by international ships of opportunity passing through the region during summer and early-fall. Here we present results from an analysis of these sections to determine the seasonal evolution of the water masses and to investigate the nature of the mesoscale variability. The mean state shows the clear presence of six water masses present at various times through the summer. The seasonal evolution of these summer water masses is characterized both in depth space and in temperature-salinity (T-S) space. Clear patterns emerge, including the arrival of Alaskan coastal water and its modification in early-fall. The primary mesoscale variability is associated with wind-driven upwelling events which occur predominantly in September. The atmospheric forcing of these events is investigated as is the oceanic response.

  10. Safety analysis -- 200 Area Savannah River Plant, F-Canyon Operations. Supplement 4

    Energy Technology Data Exchange (ETDEWEB)

    Beary, M.M.; Collier, C.D.; Fairobent, L.A.; Graham, R.F.; Mason, C.L.; McDuffee, W.T.; Owen, T.L.; Walker, D.H.

    1986-02-01

    The F-Canyon facility is located in the 200 Separations Area and uses the Purex process to recover plutonium from reactor-irradiated uranium. The irradiated uranium is normally in the form of solid or hollow cylinders called slugs. These slugs are encased in aluminum cladding and are sent to the F-Canyon from the Savannah River Plant (SRP) reactor areas or from the Receiving Basin for Offsite Fuels (RBOF). This Safety Analysis Report (SAR) documents an analysis of the F-Canyon operations and is an update to a section of a previous SAR. The previous SAR documented an analysis of the entire 200 Separations Area operations. This SAR documents an analysis of the F-Canyon and is one of a series of documents for the Separations Area as specified in the Savannah River Implementation Plans. A substantial amount of the information supporting the conclusions of this SAR is found in the Systems Analysis. Some F-Canyon equipment has been updated during the time between the Systems Analysis and this SAR and a complete description of this equipment is included in this report. The primary purpose of the analysis was to demonstrate that the F-Canyon can be operated without undue risk to onsite or offsite populations and to the environment. In this report, risk is defined as the expected frequency of an accident, multiplied by the resulting radiological consequence in person-rem. The units of risk for radiological dose are person-rem/year. Maximum individual exposure values have also been calculated and reported.

  11. Solar coal gasification reactor with pyrolysis gas recycle

    Science.gov (United States)

    Aiman, William R.; Gregg, David W.

    1983-01-01

    Coal (or other carbonaceous matter, such as biomass) is converted into a duct gas that is substantially free from hydrocarbons. The coal is fed into a solar reactor (10), and solar energy (20) is directed into the reactor onto coal char, creating a gasification front (16) and a pyrolysis front (12). A gasification zone (32) is produced well above the coal level within the reactor. A pyrolysis zone (34) is produced immediately above the coal level. Steam (18), injected into the reactor adjacent to the gasification zone (32), reacts with char to generate product gases. Solar energy supplies the energy for the endothermic steam-char reaction. The hot product gases (38) flow from the gasification zone (32) to the pyrolysis zone (34) to generate hot char. Gases (38) are withdrawn from the pyrolysis zone (34) and reinjected into the region of the reactor adjacent the gasification zone (32). This eliminates hydrocarbons in the gas by steam reformation on the hot char. The product gas (14) is withdrawn from a region of the reactor between the gasification zone (32) and the pyrolysis zone (34). The product gas will be free of tar and other hydrocarbons, and thus be suitable for use in many processes.

  12. Gas-cooled breeder reactor safety

    Energy Technology Data Exchange (ETDEWEB)

    Chermanne, J.; Burgsmueller, P. [Societe Belge pour l' Industrie Nucleaire, Brussels

    1981-01-15

    The European Association for the Gas-cooled Breeder Reactor (G B R A), set-up in 1969 prepared between 1972 and 1974 a 1200 MWe Gas-cooled Breeder Reactor (G B R) commercial reference design G B R 4. It was then found necessary that a sound and neutral appraisal of the G B R licenseability be carried out. The Commission of the European Communities (C E C) accepted to sponsor this exercise. At the beginning of 1974, the C E C convened a group of experts to examine on a Community level, the safety documents prepared by the G B R A. A working party was set-up for that purpose. The experts examined a ''Preliminary Safety Working Document'' on which written questions and comments were presented. A ''Supplement'' containing the answers to all the questions plus a detailed fault tree and reliability analysis was then prepared. After a final study of this document and a last series of discussions with G B R A representatives, the experts concluded that on the basis of the evidence presented to the Working Party, no fundamental reasons were identified which would prevent a Gas-cooled Breeder Reactor of the kind proposed by the G B R A achieving a satisfactory safety status. Further work carried out on ultimate accident have confirmed this conclusion. One can therefore claim that the overall safety risk associated with G B R s compares favourably with that of any other reactor system.

  13. NOVEL REACTOR FOR THE PRODUCTION OF SYNTHESIS GAS

    Energy Technology Data Exchange (ETDEWEB)

    Vasilis Papavassiliou; Leo Bonnell; Dion Vlachos

    2004-12-01

    Praxair investigated an advanced technology for producing synthesis gas from natural gas and oxygen This production process combined the use of a short-reaction time catalyst with Praxair's gas mixing technology to provide a novel reactor system. The program achieved all of the milestones contained in the development plan for Phase I. We were able to develop a reactor configuration that was able to operate at high pressures (up to 19atm). This new reactor technology was used as the basis for a new process for the conversion of natural gas to liquid products (Gas to Liquids or GTL). Economic analysis indicated that the new process could provide a 8-10% cost advantage over conventional technology. The economic prediction although favorable was not encouraging enough for a high risk program like this. Praxair decided to terminate development.

  14. Gas turbine modular helium reactor in cogeneration; Turbina de gas reactor modular con helio en cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Leon de los Santos, G. [UNAM, Facultad de Ingenieria, Division de Ingenieria Electrica, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Mexico, D. F. (Mexico)], e-mail: tesgleon@gmail.com

    2009-10-15

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  15. An analysis of the falling film gas-liquid reactor

    NARCIS (Netherlands)

    Davis, E.J.; Ouwerkerk-Dijkers, van M.P.; Venkatesh, S.

    1979-01-01

    A mathematical model of the falling film reactor is developed to predict the conversion and temperature distribution in the reactor as a function of the gas and liquid flow rates, physical properties, the feed composition of the reactive gas and carrier gas and other parameters of the system.

  16. Main gas circulator for VG-400 reactor plant

    International Nuclear Information System (INIS)

    Mitenkov, F.M.; Kostin, V.I.; Novinskij, E.G.; Kuropatov, A.I.; Protsenko, A.N.; Smirnov, V.P.; Stolyarevskij, A.Ya.

    1988-01-01

    Principle parameters and operating conditions of the main gas circulator (MGC) in VG-400 reactor plant are presented. Brief MGC design description and experimental work scope are given. (author). 4 refs, 4 figs, 1 tab

  17. Overview of gas cooled reactors' applications with CATHARE

    International Nuclear Information System (INIS)

    Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

    2005-01-01

    Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

  18. Analysis of Radioactivity Contamination Level of Kartini Reactor Efluen Gas to the Environment

    International Nuclear Information System (INIS)

    Suratman; Purwanto; Aminjoyo, S

    1996-01-01

    The analysis of radioactivity contamination level of Kartini reactor efluen gas to the environment has been done from 13-10-'95 until 8-2-'96. The aim of this research is to determine the radioactivity contamination level on the environment resulted from the release of Kartini reactor efluen gas and other facilities at Yogyakarta Nuclear Research Centre through stack. The analysis methods is the student t-test, the first count factor test and the gamma spectrometry. The gas sampling were carried out in the stack reactor, reactor room, environment and in other room for comparison. Efluen gas was sucked through a filter by a high volume vacuum pump. The filter was counted for beta, gamma and alpha activities. The radioactivity contamination level of the efluen gas passing through the stack to the environment was measured between 0.57 - 1.34 Bq/m3, which was equal to the airborne radioactivity in environment between 0.69 - 1.12 Bq/m3. This radioactivity comes from radon daughter, decay products result from the natural uranium and thorium series of the materials of the building

  19. Comparison of Direct and Indirect Gas Reactor Brayton Systems for Nuclear Electric Space Propulsion

    International Nuclear Information System (INIS)

    M Postlehwait; P DiLorenzo; S Belanger; J Ashcroft

    2005-01-01

    Gas reactor systems are being considered as candidates for use in generating power for the Prometheus-1 spacecraft, along with other NASA missions as part of the Prometheus program. Gas reactors offer a benign coolant, which increases core and structural materials options. However, the gas coolant has inferior thermal transport properties, relative to other coolant candidates such as liquid metals. This leads to concerns for providing effective heat transfer and for minimizing pressure drop within the reactor core. In direct gas Brayton systems, i.e. those with one or more Brayton turbines in the reactor cooling loop, the ability to provide effective core cooling and low pressure drop is further constrained by the need for a low pressure, high molecular weight gas, typically a mixture of helium and xenon. Use of separate primary and secondary gas loops, one for the reactor and one or more for the Brayton system(s) separated by heat exchanger(s), allows for independent optimization of the pressure and gas composition of each loop. The reactor loop can use higher pressure pure helium, which provides improved heat transfer and heat transport properties, while the Brayton loop can utilize lower pressure He-Xe. However, this approach requires a separate primary gas circulator and also requires gas to gas heat exchangers. This paper focuses on the trade-offs between the direct gas reactor Brayton system and the indirect gas Brayton system. It discusses heat exchanger arrangement and materials options and projects heat exchanger mass based on heat transfer area and structural design needs. Analysis indicates that these heat exchangers add considerable mass, but result in reactor cooling and system resiliency improvements

  20. CEA programme on gas cooled reactors

    International Nuclear Information System (INIS)

    Carre, F.; Fiorini, G.L.; Chapelot, Ph.; Gauthier, J.C.

    2002-01-01

    Future nuclear energy systems studies conducted by the CEA aim at investigating and developing promising technologies for future reactors, fuels and fuel cycles, for nuclear power to play a major part in sustainable energy policies. Reactors and fuel cycles are considered as integral parts of a nuclear system to be optimised as a whole. Major goals assigned to future nuclear energy systems are the following: reinforced economic competitiveness with other electricity generation means, with a special emphasis on reducing the investment cost; enhanced reliability and safety, through an improved management of reactor operation in normal and abnormal plant conditions; minimum production of long lived radioactive waste; resource saving through an effective and flexible use of the available resources of fissile and fertile materials; enhanced resistance to proliferation risks. The three latter goals are essential for the sustainability of nuclear energy in the long term. Additional considerations such as the potentialities for other applications than electricity generation (co-generation, production of hydrogen, sea water desalination) take on an increasing importance. Sustainability goals call for fast neutron spectra (to transmute nuclear waste and to breed fertile fuel) and for recycling actinides from the spent fuel (plutonium and minor actinides). New applications and economic competitiveness call for high temperature technologies (850 deg C), that afford high conversion efficiencies and hence less radioactive waste production and discharged heat. These orientations call for breakthroughs beyond light water reactors. Therefore, as a result of a screening review of candidate technologies, the CEA has selected an innovative concept of high temperature gas cooled reactor with a fast neutron spectrum, robust refractory fuel, direct conversion with a gas turbine, and integrated on-site fuel cycle as a promising system for a sustainable energy development. This objective

  1. Design and development of gas turbine high temperature reactor 300

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Yan, Xing; Takizuka, Takakazu

    2003-01-01

    JAERI (Japan Atomic Energy Research Institute) has been designing a Japan's original gas turbine high temperature reactor, GTHTR300 (Gas Turbine High Temperature Reactor 300). The greatly simplified design based on salient features of the HTGR (High Temperature Gas-cooled reactor) with a closed helium gas turbine enables the GTHTR300 a high efficient and economically competitive reactor to be deployed in early 2010s. Also, the GTHTR300 fully taking advantage of various experiences accumulated in design, construction and operation of the HTTR (High Temperature Engineering Test Reactor) and fossil gas turbine systems reduces technological development concerning a reactor system and electric generation system. Original features of this system are core design with two-year refueling interval, conventional steel material usage for a reactor pressure vessel, innovative plant flow scheme and horizontally installed gas turbine unit. Due to these salient features, the capital cost of the GTHTR300 is less than a target cost of 200 thousands Yen/kWe, and the electric generation cost is close to a target cost of 4 Yen/kWh. This paper describes the original design features focusing on reactor core design, fuel design, in-core structure design and reactor pressure vessel design except PCU design. Also, R and D for developing the power conversion unit is briefly described. The present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  2. Carbon transport in Monterey Submarine Canyon

    Science.gov (United States)

    Barry, J.; Paull, C. K.; Xu, J. P.; Clare, M. A.; Gales, J. A.; Buck, K. R.; Lovera, C.; Gwiazda, R.; Maier, K. L.; McGann, M.; Parsons, D. R.; Simmons, S.; Rosenberger, K. J.; Talling, P. J.

    2017-12-01

    Submarine canyons are important conduits for sediment transport from continental margins to the abyss, but the rate, volume, and time scales of material transport have been measured only rarely. Using moorings with current meters, sediment traps (10 m above bottom) and optical backscatter sensors, we measured near-bottom currents, suspended sediment concentrations, and sediment properties at 1300 m depth in Monterey Canyon and at a non-canyon location on the continental slope at the same depth. Flow and water column backscatter were used to characterize "ambient" conditions when tidal currents dominated the flow field, and occasional "sediment transport events" when anomalously high down-canyon flow with sediment-laden waters arrived at the canyon mooring. The ambient sediment flux measured in sediment traps in Monterey Canyon was 350 times greater than measured at the non-canyon location. Although the organic carbon content of the canyon sediment flux during ambient periods was low (1.8 %C) compared to the slope location (4.9 %C), the ambient carbon transport in the canyon was 130 times greater than at the non-canyon site. Material fluxes during sediment transport events were difficult to measure owing to clogging of sediment traps, but minimal estimates indicate that mass transport during events exceeds ambient sediment fluxes through the canyon by nearly 3 orders of magnitude, while carbon transport is 380 times greater. Estimates of the instantaneous and cumulative flux of sediment and carbon from currents, backscatter, and sediment properties indicated that: 1) net flux is down-canyon, 2) flux is dominated by sediment transport events, and 3) organic carbon flux through 1300 m in Monterey Canyon was ca. 1500 MT C per year. The injection of 1500 MTCy-1 into the deep-sea represents ca. 260 km2 of the sediment C flux measured at the continental slope station (5.8 gCm-2y-1) and is sufficient to support a benthic community carbon demand of 5 gCm-2y-1 over 300 km2.

  3. Gas-Cooled Reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1978-01-01

    Gas-Cooled Reactors are considered to have a significant future impact on the application of fission energy. The specific types are the steam-cycle High-Temperature Gas-Cooled Reactor, the Gas-Cooled Fast Breeder Reactor, the gas-turbine HTGR, and the Very High-Temperature Process Heat Reactor. The importance of developing the above systems is discussed relative to alternative fission power systems involving Light Water Reactors, Heavy Water Reactors, Spectral Shift Controlled Reactors, and Liquid-Metal-Cooled Fast Breeder Reactors. A primary advantage of developing GCRs as a class lies in the technology and cost interrelations, permitting cost-effective development of systems having diverse applications. Further, HTGR-type systems have highly proliferation-resistant characteristics and very attractive safety features. Finally, such systems and GCFRs are mutally complementary. Overall, GCRs provide interrelated systems that serve different purposes and needs; their development can proceed in stages that provide early benefits while contributing to future needs. It is concluded that the long-term importance of the various GCRs is as follows: HTGR, providing a technology for economic GCFRs and HTGR-GTs, while providing a proliferation-resistant reactor system having early economic and fuel utilization benefits; GCFR, providing relatively low cost fissile fuel and reducing overall separative work needs at capital costs lower than those for LMFBRs; HTGR-GT (in combination with a bottoming cycle), providing a very high thermal efficiency system having low capital costs and improved fuel utilization and technology pertinent to VHTRs; HTGR-GT, providing a power system well suited for dry cooling conditions for low-temperature process heat needs; and VHTR, providing a high-temperature heat source for hydrogen production processes

  4. Gas-phase photocatalysis in μ-reactors

    DEFF Research Database (Denmark)

    Vesborg, Peter Christian Kjærgaard; Olsen, Jakob Lind; Henriksen, Toke Riishøj

    2010-01-01

    Gas-phase photocatalysis experiments may benefit from the high sensitivity and good time response in product detection offered by μ-reactors. We demonstrate this by carrying out CO oxidation and methanol oxidation over commercial TiO2 photocatalysts in our recently developed high-sensitivity reac......Gas-phase photocatalysis experiments may benefit from the high sensitivity and good time response in product detection offered by μ-reactors. We demonstrate this by carrying out CO oxidation and methanol oxidation over commercial TiO2 photocatalysts in our recently developed high...

  5. Gas-liquid reactor / separator: dynamics and operability characteristics

    NARCIS (Netherlands)

    Ranade, V.; Kuipers, J.A.M.; Versteeg, Geert

    1999-01-01

    A comprehensive mathematical model is developed to simulate gas¿liquid reactor in which both, reactants as well as products enter or leave the reactor in gas phase while the reactions take place in liquid phase. A case of first-order reaction (isothermal) was investigated in detail using the dynamic

  6. High temperature gas cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hosegood, S.B.; Lockett, G.E.

    1975-01-01

    For high-temperature gas cooled reactors it is considered advantageous to design the core so that the moderator blocks can be removed and replaced by some means of standpipes normally situated in the top of the reactor vessel. An arrangement is here described to facilitate these operations. The blocks have end faces shaped as irregular hexagons with three long sides of equal length and three short sides also of equal length, one short side being located between each pair of adjacent long sides, and the long sides being inclined towards one another at 60 0 . The block defines a number of coolant channels located parallel to its sides. Application of the arrangement to a high temperature gas-cooled reactor with refuelling standpipes is described. The standpipes are located in the top of the reactor vessel above the tops of the columns and are disposed coaxially above the hexagonal channels, with diameters that allow the passage of the blocks. (U.K.)

  7. Modelling of non-catalytic reactors in a gas-solid trickle flow reactor: Dry, regenerative flue gas desulphurization using a silica-supported copper oxide sorbent

    NARCIS (Netherlands)

    Kiel, J.H.A.; Kiel, J.H.A.; Prins, W.; van Swaaij, Willibrordus Petrus Maria

    1992-01-01

    A one-dimensional, two-phase dispersed plug flow model has been developed to describe the steady-state performance of a relatively new type of reactor, the gas-solid trickle flow reactor (GSTFR). In this reactor, an upward-flowing gas phase is contacted with as downward-flowing dilute solids phase

  8. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics.

  9. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    International Nuclear Information System (INIS)

    1988-01-01

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

  10. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  11. Axial and Radial Gas Holdup in Bubble Column Reactor

    International Nuclear Information System (INIS)

    Wagh, Sameer M.; Ansari, Mohashin E Alan; Kene, Pragati T.

    2014-01-01

    Bubble column reactors are considered the reactor of choice for numerous applications including oxidation, hydrogenation, waste water treatment, and Fischer-Tropsch (FT) synthesis. They are widely used in a variety of industrial applications for carrying out gas-liquid and gas-liquid-solid reactions. In this paper, the computational fluid dynamics (CFD) model is used for predicting the gas holdup and its distribution along radial and axial direction are presented. Gas holdup increases linearly with increase in gas velocity. Gas bubbles tends to concentrate more towards the center of the column and follows a wavy path

  12. Study on gas-liquid loop reactors with annular bubbling

    International Nuclear Information System (INIS)

    Fei, L.M.; Wang, S.X.; Wu, X.Q.; Lu, D.W.

    1987-01-01

    Bubbling column with draft tube is one of nearly developed reactor. On the background of hydrocarbon oxidations and biochemical engineerings, it has been widely used in chemical industry due to the well characteristics of mass and heat transfer. In this paper, the characteristics of fluid flow, such as gas hold-up, backmixing and mass transfer referred to the liquid volume were measured in a gas-liquid loop reactor with annular bubbling. Different materials - water, alcohol and oi l- were used in the study in measuring the gas hold-up in the annular of the reactor

  13. A study of silver behavior in Gas-turbine High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Sawa, Kazuhiro; Tanaka, Toshiyuki

    1995-11-01

    A Gas-turbine High Temperature Gas-cooled Reactor (GT-HTGR) is one of the promising reactor systems of future HTGRs. In the design of GT-HTGR, behavior of fission products, especially of silver, is considered to be important from the view point of maintenance of gas-turbine. A study of silver behavior in the GT-HTGR was carried out based on current knowledge. The purposes of this study were to determine an importance of the silver problem quantitatively, countermeasures to the problem and items of future research and development which will be needed. In this study, inventory, fractional release from fuel, plateout in the primary circuit and radiation dose were evaluated, respectively. Based on this study, it is predicted that gamma-ray from plateout silver in gas-turbine system contributes about a half of total radiation dose after reactor shutdown. In future, more detail data for silver release from fuel, plateout behavior, etc. using the High Temperature Engineering Test Reactor (HTTR), for example, will be needed to carry out reasonable design. (author)

  14. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  15. Gas export potential of Russia's East: Will it match Asia-Pacific markets?

    International Nuclear Information System (INIS)

    Khartukov, E.; Starostina, E.

    2002-01-01

    Russia's Far East and East Siberia are emerging as new major sources of gas supplies for East Asian energy markets. Thanks to ongoing and earmarked resource and infrastructure developments in Sakhalin, Yakutia (Sakha) and Irkutsk, by around 2020 these poorly developed but naturally endowed areas of the country's East can provide between 50 and 70 Bcm/yr (5-7 Bcfd) of natural gas, including up to 10 Mt/yr of LNG, available for exports to neighbouring Pacific countries (primarily to the PRC, Japan, South Korea as well as to Taiwan and the U.S. West Coast). This can noticeably reshape today's matrix of the Asia-Pacific energy flows and even destabilize the regional gas market. (author)

  16. Measurement of dissolved hydrogen and hydrogen gas transfer in a hydrogen-producing reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shizas, I.; Bagley, D.M. [Toronto Univ., ON (Canada). Dept. of Civil Engineering

    2004-07-01

    This paper presents a simple method to measure dissolved hydrogen concentrations in the laboratory using standard equipment and a series of hydrogen gas transfer tests. The method was validated by measuring hydrogen gas transfer parameters for an anaerobic reactor system that was purged with 10 per cent carbon dioxide and 90 per cent nitrogen using a coarse bubble diffuser stone. Liquid samples from the reactor were injected into vials and hydrogen was allowed to partition between the liquid and gaseous phases. The concentration of dissolved hydrogen was determined by comparing the headspace injections onto a gas chromatograph and a standard curve. The detection limit was 1.0 x 10{sup -5} mol/L of dissolved hydrogen. The gas transfer rate for hydrogen in basal medium and anaerobic digester sludge was used to validate the method. Results were compared with gas transfer models. In addition to monitoring dissolved hydrogen in reactor systems, this method can help improve hydrogen production potential. 1 ref., 4 figs.

  17. Fission gas behaviour in water reactor fuels

    International Nuclear Information System (INIS)

    2002-01-01

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

  18. Gas core reactors for coal gasification

    International Nuclear Information System (INIS)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H 2 and CO in the reactor cavity, indicating a 98 percent conversion of water and coal at only 1500 0 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H 2 O to CO 2 and H 2 . Furthermore, it is shown the H 2 obtained per pound of carbon has 23 percent greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H 2 , fresh water and sea salts from coal

  19. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  20. Zeolite Membrane Reactor for Water Gas Shift Reaction for Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Jerry Y.S. [Arizona State Univ., Mesa, AZ (United States)

    2013-01-29

    Gasification of biomass or heavy feedstock to produce hydrogen fuel gas using current technology is costly and energy-intensive. The technology includes water gas shift reaction in two or more reactor stages with inter-cooling to maximize conversion for a given catalyst volume. This project is focused on developing a membrane reactor for efficient conversion of water gas shift reaction to produce a hydrogen stream as a fuel and a carbon dioxide stream suitable for sequestration. The project was focused on synthesizing stable, hydrogen perm-selective MFI zeolite membranes for high temperature hydrogen separation; fabricating tubular MFI zeolite membrane reactor and stable water gas shift catalyst for membrane reactor applications, and identifying experimental conditions for water gas shift reaction in the zeolite membrane reactor that will produce a high purity hydrogen stream. The project has improved understanding of zeolite membrane synthesis, high temperature gas diffusion and separation mechanisms for zeolite membranes, synthesis and properties of sulfur resistant catalysts, fabrication and structure optimization of membrane supports, and fundamentals of coupling reaction with separation in zeolite membrane reactor for water gas shift reaction. Through the fundamental study, the research teams have developed MFI zeolite membranes with good perm-selectivity for hydrogen over carbon dioxide, carbon monoxide and water vapor, and high stability for operation in syngas mixture containing 500 part per million hydrogen sulfide at high temperatures around 500°C. The research teams also developed a sulfur resistant catalyst for water gas shift reaction. Modeling and experimental studies on the zeolite membrane reactor for water gas shift reaction have demonstrated the effective use of the zeolite membrane reactor for production of high purity hydrogen stream.

  1. Gas-cooled reactors: the importance of their development

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U 3 O 8 before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production

  2. Evaluation of tritium production rate in a gas-cooled reactor with continuous tritium recovery system for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Matsuura, Hideaki, E-mail: mat@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Nakaya, Hiroyuki; Nakao, Yasuyuki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 819-0395 (Japan); Shimakawa, Satoshi; Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki 311-1393 (Japan); Nishikawa, Masabumi [Graduate School of Engineering Science, Kyushu University, 6-10-1 Hakozaki, Fukuoka 812-8581 (Japan)

    2013-10-15

    Highlights: • The performance of a gas-cooled reactor as a tritium production system was studied. • A continuous tritium recovery using helium gas was considered. • Gas-cooled reactors with 3 GW output in all can produce ∼6 kg of tritium in a year • Performance of the system was examined for Li{sub 4}SiO{sub 4}, Li{sub 2}TiO{sub 3} and LiAlO{sub 2} compounds. -- Abstract: The performance of a high-temperature gas-cooled reactor as a tritium production with continuous tritium recovery system is examined. A gas turbine high-temperature reactor of 300-MWe (600 MW) nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations for the three-dimensional entire-core region of the GTHTR300 were performed. A Li loading pattern for the continuous tritium recovery system in the gas-cooled reactor is presented. It is shown that module gas-cooled reactors with a total thermal output power of 3 GW in all can produce ∼6 kg of tritium maximum in a year.

  3. Heavy water moderated gas-cooled reactors

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  4. Photography and Neobaroque Imaginary in Julio Cortázar's "Las babas del diablo": Can the Neobaroque Name a Photograph?

    Science.gov (United States)

    Hakobyan, Liana

    2018-01-01

    This article examines Julio Cortázar's short story "Las babas del diablo" from a visual perspective and at the intersection of Roland Barthes's ideas on photography and Severo Sarduy's theory on the Neobaroque. I propose that in "Las babas del diablo" photography and the Neobaroque--two seemingly unrelated concepts--interact…

  5. Late Holocene earthquake history of the Brigham City segment of the Wasatch fault zone at the Hansen Canyon, Kotter Canyon, and Pearsons Canyon trench sites, Box Elder County, Utah

    Science.gov (United States)

    DuRoss, Christopher B.; Personius, Stephen F.; Crone, Anthony J.; McDonald, Greg N.; Briggs, Richard W.

    2012-01-01

    Of the five central segments of the Wasatch fault zone (WFZ) having evidence of recurrent Holocene surface-faulting earthquakes, the Brigham City segment (BCS) has the longest elapsed time since its most recent surface-faulting event (~2.1 kyr) compared to its mean recurrence time between events (~1.3 kyr). Thus, the BCS has the highest time-dependent earthquake probability of the central WFZ. We excavated trenches at three sites––the Kotter Canyon and Hansen Canyon sites on the north-central BCS and Pearsons Canyon site on the southern BCS––to determine whether a surface-faulting earthquake younger than 2.1 ka occurred on the BCS. Paleoseismic data for Hansen Canyon and Kotter Canyon confirm that the youngest earthquake on the north-central BCS occurred before 2 ka, consistent with previous north-central BCS investigations at Bowden Canyon and Box Elder Canyon. At Hansen Canyon, the most recent earthquake is constrained to 2.1–4.2 ka and had 0.6–2.5 m of vertical displacement. At Kotter Canyon, we found evidence for two events at 2.5 ± 0.3 ka and 3.5 ± 0.3 ka, with an average displacement per event of 1.9–2.3 m. Paleoseismic data from Pearsons Canyon, on the previously unstudied southern BCS, indicate that a post-2 ka earthquake ruptured this part of the segment. The Pearsons Canyon earthquake occurred at 1.2 ± 0.04 ka and had 0.1–0.8 m of vertical displacement, consistent with our observation of continuous, youthful scarps on the southern 9 km of the BCS having 1–2 m of late Holocene(?) surface offset. The 1.2-ka earthquake on the southern BCS likely represents rupture across the Weber–Brigham City segment boundary from the penultimate Weber-segment earthquake at about 1.1 ka. The Pearsons Canyon data result in a revised length of the BCS that has not ruptured since 2 ka (with time-dependent probability implications), and provide compelling evidence of at least one segment-boundary failure and multi-segment rupture on the central WFZ. Our

  6. The impact of ion exchange media and filters on LLW processing

    International Nuclear Information System (INIS)

    James, K.L.; Miller, C.C.

    1992-01-01

    Optimized ion exchange media at Diablo Canyon have steadily improved the treatment of radioactive liquid waste. The activity released to the environment has been reduced while simultaneously reducing the volume of solid radwaste generated from processing radioactive liquids. This has lowered the liquid waste processing costs and reduced the number of radioactive shipments from the plant. A cobalt treatment technique was identified and successfully implemented prior to reactor coolant chemistry alteration. A cesium treatment using zeolite has been successfully implemented. A cobalt removal treatment, combining series cation ion exchange with submicron filtration, has successfully removed cobalt after reactor coolant chemistry alteration. A new carbon-based material will be monitored to find a media to remove cobalt from high-conductivity liquids. (author)

  7. Overview of environmental control aspects for the gas-cooled fast reactor

    International Nuclear Information System (INIS)

    Nolan, A.M.

    1981-05-01

    Environmental control aspects relating to release of radionuclides have been analyzed for the Gas-Cooled Fast Reactor (GCFR). Information on environmental control systems was obtained for the most recent GCFR designs, and was used to evaluate the adequacy of these systems. The GCFR has been designed by the General Atomic Company as an alternative to other fast breeder reactor designs, such as the Liquid Metal Fast Breeder Reactor (LMFBR). The GCFR design includes mixed oxide fuel and helium coolant. The environmental impact of expected radionuclide releases from normal operation of the GCFR was evaluated using estimated collective dose equivalent commitments resulting from 1 year of plant operation. The results were compared to equivalent estimates for the Light Water Reactor (LWR) and High-Temperature Gas-Cooled Reactor (HTGR). A discussion of uncertainties in system performances, tritium production rates, and radiation quality factors for tritium is included

  8. High-resolution seismic imaging of the gas and gas hydrate system at Green Canyon 955 in the Gulf of Mexico

    Science.gov (United States)

    Haines, S. S.; Hart, P. E.; Collett, T. S.; Shedd, W. W.; Frye, M.

    2015-12-01

    High-resolution 2D seismic data acquired by the USGS in 2013 enable detailed characterization of the gas and gas hydrate system at lease block Green Canyon 955 (GC955) in the Gulf of Mexico, USA. Earlier studies, based on conventional industry 3D seismic data and logging-while-drilling (LWD) borehole data acquired in 2009, identified general aspects of the regional and local depositional setting along with two gas hydrate-bearing sand reservoirs and one layer containing fracture-filling gas hydrate within fine-grained sediments. These studies also highlighted a number of critical remaining questions. The 2013 high-resolution 2D data fill a significant gap in our previous understanding of the site by enabling interpretation of the complex system of faults and gas chimneys that provide conduits for gas flow and thus control the gas hydrate distribution observed in the LWD data. In addition, we have improved our understanding of the main channel/levee sand reservoir body, mapping in fine detail the levee sequences and the fault system that segments them into individual reservoirs. The 2013 data provide a rarely available high-resolution view of a levee reservoir package, with sequential levee deposits clearly imaged. Further, we can calculate the total gas hydrate resource present in the main reservoir body, refining earlier estimates. Based on the 2013 seismic data and assumptions derived from the LWD data, we estimate an in-place volume of 840 million cubic meters or 29 billion cubic feet of gas in the form of gas hydrate. Together, these interpretations provide a significantly improved understanding of the gas hydrate reservoirs and the gas migration system at GC955.

  9. 76 FR 35208 - Pacific Gas and Electric Company; Nevada Irrigation District; Notice of Environmental Site Review

    Science.gov (United States)

    2011-06-16

    ... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [ Project No. 2310-193--California; Project No. 2266-102--California] Pacific Gas and Electric Company; Nevada Irrigation District; Notice of Environmental Site Review On July 6-8, 2011, the Federal Energy Regulatory Commission (Commission) staff and the Pacific Gas and Electric Company ...

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

    International Nuclear Information System (INIS)

    1989-06-01

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

  11. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies

  12. Gas pollutant cleaning by a membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Topis, S.; Koutsonikolas, D.; Kaldis, S. (and others) [Aristotle University of Thessaloniki, Thessaloniki (Greece). Dept. of Chemical Engineering

    2005-07-01

    An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al{sub 2}O{sub 3} catalyst was prepared by the dry and wet impregnation method and characterized by ICP, SEM, XRD and N{sub 2} adsorption before and after activation. Commercially available {alpha}-Al{sub 2}O{sub 3} membranes were also characterized and the permeabilities and selectivities of H{sub 2}, N{sub 2} and CO{sub 2} were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. 5 refs., 6 figs., 1 tab.

  13. Gas pollutant cleaning by a membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    George E. Skodras; Sotiris Kaldis; Savas G. Topis; Dimitris Koutsonikolas; George P. Sakellaropoulos [Aristotle University of Thessaloniki, Thessaloniki (Greece). Chemical Process Engineering Laboratory, Dept. of Chemical Engineering

    2006-07-01

    An alternative technology for the removal of gas pollutants at the intergrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al{sub 2}O{sub 3} catalyst was prepared by the dry and wet impregnation method and characterized by ICP, SEM, XRD and N{sub 2} adsorption before and after activation. Commercially available {alpha}-Al{sub 2}O{sub 3} membranes were also characterized and the permeabilities and permselectivities of H{sub 2}, N{sub 2} and CO{sub 2} were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. 9 refs., 6 figs., 1 tab.

  14. Operational Readiness Review Final Report For F-Canyon Restart. Phase 1

    Energy Technology Data Exchange (ETDEWEB)

    McFarlane, A.F.; Spangler, J.B.

    1995-04-05

    An independent WSRC Operational Readiness Review was performed for the restart of Phase 1 processing in F-Canyon, Building 221-F. Readiness to restart the Second Plutonium Cycle process and solvent recovery was assessed. The ORR was conducted by an ORR board of ten members with the support of a subject matter expert. The chairman and four members were drawn from the Operational Safety Evaluation Department, ESH& QA Division; additional members were drawn from other WSRC divisions, independent of the F-Canyon operating division (NMPD). Based on the results of the readiness verification assessments performed according to the ORR plan and the validation of pre-restart corrective actions, the WSRC independent ORR Board has concluded that the facility has achieved the state of readiness committed to in the Restart Plan. Also, based on the scope of the ORR, it is the opinion of the board that F-Canyon Phase 1 processes can be restarted without undue risk to the safety of the public and onsite workers and without undue risk to the environment.

  15. Operational Readiness Review Final Report For F-Canyon Restart. Phase 1

    International Nuclear Information System (INIS)

    McFarlane, A.F.; Spangler, J.B.

    1995-01-01

    An independent WSRC Operational Readiness Review was performed for the restart of Phase 1 processing in F-Canyon, Building 221-F. Readiness to restart the Second Plutonium Cycle process and solvent recovery was assessed. The ORR was conducted by an ORR board of ten members with the support of a subject matter expert. The chairman and four members were drawn from the Operational Safety Evaluation Department, ESH ampersand QA Division; additional members were drawn from other WSRC divisions, independent of the F-Canyon operating division (NMPD). Based on the results of the readiness verification assessments performed according to the ORR plan and the validation of pre-restart corrective actions, the WSRC independent ORR Board has concluded that the facility has achieved the state of readiness committed to in the Restart Plan. Also, based on the scope of the ORR, it is the opinion of the board that F-Canyon Phase 1 processes can be restarted without undue risk to the safety of the public and onsite workers and without undue risk to the environment

  16. The Whittard Canyon - A case study of submarine canyon processes

    Science.gov (United States)

    Amaro, T.; Huvenne, V. A. I.; Allcock, A. L.; Aslam, T.; Davies, J. S.; Danovaro, R.; De Stigter, H. C.; Duineveld, G. C. A.; Gambi, C.; Gooday, A. J.; Gunton, L. M.; Hall, R.; Howell, K. L.; Ingels, J.; Kiriakoulakis, K.; Kershaw, C. E.; Lavaleye, M. S. S.; Robert, K.; Stewart, H.; Van Rooij, D.; White, M.; Wilson, A. M.

    2016-08-01

    Submarine canyons are large geomorphological features that incise continental shelves and slopes around the world. They are often suggested to be biodiversity and biomass hotspots, although there is no consensus about this in the literature. Nevertheless, many canyons do host diverse faunal communities but owing to our lack of understanding of the processes shaping and driving this diversity, appropriate management strategies have yet to be developed. Here, we integrate all the current knowledge of one single system, the Whittard Canyon (Celtic Margin, NE Atlantic), including the latest research on its geology, sedimentology, geomorphology, oceanography, ecology, and biodiversity in order to address this issue. The Whittard Canyon is an active system in terms of sediment transport. The net suspended sediment transport is mainly up-canyon causing sedimentary overflow in some upper canyon areas. Occasionally sediment gravity flow events do occur, some possibly the result of anthropogenic activity. However, the role of these intermittent gravity flows in transferring labile organic matter to the deeper regions of the canyon appears to be limited. More likely, any labile organic matter flushed downslope in this way becomes strongly diluted with bulk material and is therefore of little food value for benthic fauna. Instead, the fresh organic matter found in the Whittard Channel mainly arrives through vertical deposition and lateral transport of phytoplankton blooms that occur in the area during spring and summer. The response of the Whittard Canyon fauna to these processes is different in different groups. Foraminiferal abundances are higher in the upper parts of the canyon and on the slope than in the lower canyon. Meiofaunal abundances in the upper and middle part of the canyon are higher than on adjacent slopes, but lower in the deepest part. Mega- and macrofauna abundances are higher in the canyon compared with the adjacent slope and are higher in the eastern than

  17. Outer continental shelf oil and gas activities. Pacific update: August 1987 - November 1989

    Energy Technology Data Exchange (ETDEWEB)

    Slitor, Douglas L.; Wiese, Jeffrey D.; Karpas, Robert M.

    1990-01-01

    This Pacific Update focuses on the geology and petroleum potential of the Central California and Washington-Oregon OCS Planning Areas. This report discusses the following topics: offshore oil and gas resources of the Pacific region; project-specific developments and status; and magnitude and timing of offshore developments. (CBS)

  18. United States natural gas markets, contracts and risks: What lessons for the European Union and Asia-Pacific natural gas markets?

    International Nuclear Information System (INIS)

    Talus, Kim

    2014-01-01

    The article examines the natural gas markets of the United States, the European Union and the Asia-Pacific region and their regulation and contractual structures. The article's main focus is on the United States natural gas markets. The European Union and Asia-Pacific markets are compared to this more developed market. By comparing the physical and ideological characteristics of, and differences between, the three main international gas markets, the article exposes the limits of regulatory and contractual transplants in this area of law and policy. Each of these markets is unique, which limits the opportunities for modelling certain market institutions on the basis of the more developed markets in the United States. This applies for both the EU and the Asia-Pacific region. - Highlights: • Differences in the physical markets impact regulation. • Regulatory transplants have risks. • The approach in energy policy should be based on “Law-in-Context” approach

  19. Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

    International Nuclear Information System (INIS)

    Scates, Dawn M.; Hartwell, John K.; Walter, John B.; Drigert, Mark W.; Harp, Jason M.

    2010-01-01

    The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.

  20. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  1. Mechanical properties data of 2-1/4Cr-1Mo steel for the experimental very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Oku, Tatsuo; Kikuyama, Toshihiko; Fukaya, Kiyoshi; Kodaira, Tsuneo

    1978-11-01

    This is a collection of mechanical properties data of 2-1/4Cr-1Mo steel necessary for structural design and safety analysis of the pressure vessel of the Experimental Very High Temperature Gas-Cooled Reactor (VHTR). These include physical properties, mechanical properties, temper embrittlement, creep with fatigue, fracture toughness and irradiation effects. A review of the data shows the research areas to be carried out particularly in the future for more data. (author)

  2. Heat transfer tests of ribbed surfaces for gas-cooled reactors

    International Nuclear Information System (INIS)

    Klepper, O.H.

    1975-07-01

    The performance of gas-cooled reactors is often limited by the heat transfer in the reactor core. Means for modifying core heat transfer surfaces to improve their performance were investigated. The 0.3-in.-OD stainless steel clad heater rods were photo-etched to produce external ribs 0.006 in. high and 0.12 in. wide with a pitch of 0.072 in. Helical ribs with a helix angle of 37 0 (to promote interchannel flow mixing in a multirod array) were provided on one surface. For comparison purposes, a transversely ribbed surface and a smooth rod were also studied. The test surfaces were 49 in. long with a 24-in. heated region, concentrically arranged inside a smooth 0.602-in.-ID stainless steel tube. Nitrogen gas at pressures up to 400 psig was used as the coolant; the linear heat rating ranged to 6.8 kW/ft at surface temperatures up to 1400 0 F; T/sub w/T/sub b/ varied from 1.2 to 2.4 at Re values up to 450,000. Annulus results were recalculated for rod geometry using two different transformations. Good agreement was observed with applicable literature values. The effectiveness of the surfaces was assessed as the ratio E of the heat transfer coefficients of the roughened rods to that of a smooth rod at the same pumping power. The effectiveness of the spiral ribs ranged from 1.3 to 1.4, and from 1.2 to 1.4 for the transverse ribs, spanning Re values from 60,000 to 400,000. These data include variations introduced by alternate transformation methods that were used to make annulus test results applicable to rod geometry. The surfaces investigated in these tests were considered for fast gas-cooled reactors; however, the range of parameters studied also applies to heat transfer from ribbed rod-type fuel elements in thermal gas-cooled reactors. (U.S.)

  3. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, R.J.; Wright, S.A.; Lenard, R.X.; Harms, G.A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars. copyright 1999 American Institute of Physics

  4. A gas-cooled reactor surface power system

    International Nuclear Information System (INIS)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars

  5. A Gas-Cooled Reactor Surface Power System

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Lenard, R.X.; Lipinski, R.J.; Wright, S.A.

    1998-11-09

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life- cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitide clad in Nb 1 %Zr, which has been extensively tested under the SP-I 00 program The fiel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fbel and stabilizing the geometty against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality cannot occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  6. H2 gas pressure calculation of FPM capsule failure at RSG-GAS reactor core

    International Nuclear Information System (INIS)

    Hastuti, Endiah Puji; Sunaryo, Geni Rina

    2002-01-01

    RSG-GAS has been irradiated FPM capsule for 236 times, one of those i.e. capsule number 228 has failure. The one of root cause of failure possibility is radiolysis reaction can be occurred in FPM capsule when it is filled with water during irradiation in the reactor core. The safety analysis of the radiolysis reaction in the capsule has been done. The oc cumulative hydrogen gas production can cause high pressure in the capsule then a mechanical damage occurred. The analysis was done at 10 MW of reactor power which equivalent with neutron flux of 0,6929 x 10 1 4 n/cm 2 sec and γ dose rate of 0,63x10 9 rad/hour. The assumption is the capsule is filled with water at maximum volume, i.e. 176.67 ml. The results of calculation showed that radiolysis reaction with γ and neutron produce hydrogen gas for nominal flow rate each are 494 atm and 19683 atm for γ and neutron radiolysis, respectively. H 2 gas pressure for 5% flow rate each are 723 atm. and 25772 atm., for γ and neutron radiolysis, respectively. The changing of the operation condition due to radiolysis together with one way valve' phenomena, can be produce hydrogen gas from water during irradiation in the reactor core and can be the one of root cause of capsule failure. This analysis recommended the FPM capsule preparation must be guaranteed no water or/and there is no possibility of water immersion in the capsule during irradiation in the core by more accurate leak test

  7. Gas turbine modular helium reactor in cogeneration

    International Nuclear Information System (INIS)

    Leon de los Santos, G.

    2009-10-01

    This work carries out the thermal evaluation from the conversion of nuclear energy to electric power and process heat, through to implement an outline gas turbine modular helium reactor in cogeneration. Modeling and simulating with software Thermo flex of Thermo flow the performance parameters, based on a nuclear power plant constituted by an helium cooled reactor and helium gas turbine with three compression stages, two of inter cooling and one regeneration stage; more four heat recovery process, generating two pressure levels of overheat vapor, a pressure level of saturated vapor and one of hot water, with energetic characteristics to be able to give supply to a very wide gamma of industrial processes. Obtaining a relationship heat electricity of 0.52 and efficiency of net cogeneration of 54.28%, 70.2 MW net electric, 36.6 MW net thermal with 35% of condensed return to 30 C; for a supplied power by reactor of 196.7 MW; and with conditions in advanced gas turbine of 850 C and 7.06 Mpa, assembly in a shaft, inter cooling and heat recovery in cogeneration. (Author)

  8. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-12-01

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  9. An atmospheric pressure flow reactor: Gas phase kinetics and mechanism in tropospheric conditions without wall effects

    Science.gov (United States)

    Koontz, Steven L.; Davis, Dennis D.; Hansen, Merrill

    1988-01-01

    A new type of gas phase flow reactor, designed to permit the study of gas phase reactions near 1 atm of pressure, is described. A general solution to the flow/diffusion/reaction equations describing reactor performance under pseudo-first-order kinetic conditions is presented along with a discussion of critical reactor parameters and reactor limitations. The results of numerical simulations of the reactions of ozone with monomethylhydrazine and hydrazine are discussed, and performance data from a prototype flow reactor are presented.

  10. Natural gas turbine topping for the iris reactor

    International Nuclear Information System (INIS)

    Oriani, L.; Lombardi, C.; Paramonov, D.

    2001-01-01

    Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

  11. Natural gas turbine topping for the iris reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oriani, L.; Lombardi, C. [Politecnico di Milano, Milan (Italy); Paramonov, D. [Westinghouse Electric Corp., LLC, Pittsburgh, PA (United States)

    2001-07-01

    Nuclear power plant designs are typically characterized by high capital and low fuel costs, while the opposite is true for fossil power generation including the natural gas-fired gas turbine combined cycle currently favored by many utilities worldwide. This paper examines potential advantages of combining nuclear and fossil (natural gas) generation options in a single plant. Technical and economic feasibility and attractiveness of a gas turbine - nuclear reactor combined cycle where gas turbine exhaust is used to superheat saturated steam produced by a low power light water reactor are examined. It is shown that in a certain range of fuel and capital costs of nuclear and fossil options, the proposed cycle offers an immediate economic advantage over stand-alone plants resulting from higher efficiency of the nuclear plant. Additionally, the gas turbine topping will result in higher fuel flexibility without the economic penalty typically associated with nuclear power. (author)

  12. Modular high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shepherd, L.R.

    1988-01-01

    The high financial risk involved in building large nuclear power reactors has been a major factor in halting investment in new plant and in bringing further technical development to a standstill. Increased public concern about the safety of nuclear plant, particularly after Chernobyl, has contributed to this stagnation. Financial and technical risk could be reduced considerably by going to small modular units, which would make it possible to build up power station capacity in small steps. Such modular plant, based on the helium-cooled high temperature reactor (HTR), offers remarkable advantages in terms of inherent safety characteristics, partly because of the relatively small size of the individual modules but more on account of the enormous thermal capacity and high temperature margins of the graphitic reactor assemblies. Assessments indicate that, in the USA, the cost of power from the modular systems would be less than that from conventional single reactor plant, up to about 600 MW(e), and only marginally greater above that level, a margin that should be offset by the shorter time required in bringing the modular units on line to earn revenue. The modular HTR would be particularly appropriate in the UK, because of the considerable British industrial background in gas-cooled reactors, and could be a suitable replacement for Magnox. The modular reactor would be particularly suited to combined heat and power schemes and would offer great potential for the eventual development of gas turbine power conversion and the production of high-temperature process heat. (author)

  13. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Shropshire, D.E.; Herring, J.S.

    2004-01-01

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  14. The temperature distribution in a gas core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C. (Interuniversitair Reactor Inst., Delft (Netherlands)); Kistemaker, J.; Boersma-Klein, W.; Vitalis, F. (FOM-Instituut voor Atoom-en Molecuulfysica, Amsterdam (Netherlands))

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author).

  15. The temperature distribution in a gas core fission reactor

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.; Dam, H. van; Kuijper, J.C.; Kistemaker, J.; Boersma-Klein, W.; Vitalis, F.

    1991-01-01

    A model is proposed for the heat transport in a nuclear reactor with gaseous fuel at high temperatures taking into account radiative and kinetic heat transfer. A derivation is given of the equation determining the temperature distribution in a gas core reactor and different numerical solution methods are discussed in detail. Results are presented of the temperature distribution. The influence of the kinetic heat transport and of dissociation of the gas molecules is shown. Also discussed is the importance of the temperature gradient at the reactor wall and its dependence on system parameters. (author)

  16. The Integration Of Process Heat Applications To High Temperature Gas Reactors

    International Nuclear Information System (INIS)

    McKellar, Michael G.

    2011-01-01

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  17. Dissolution flowsheet for high flux isotope reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-09-27

    As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U3O8) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H2. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H2) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H2 and other permanent gases in the dissolution offgas allowing the development of H2 generation rate profiles. The H2 generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

  18. Core configuration of a gas-cooled reactor as a tritium production device for fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, H., E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, H.; Nakao, Y. [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Shimakawa, S.; Goto, M.; Nakagawa, S. [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan); Nishikawa, M. [Malaysia-Japan International Institute of Technology, UTM, Kuala Lumpur 54100 (Malaysia)

    2014-05-01

    The performance of a high-temperature gas-cooled reactor as a tritium production device is examined, assuming the compound LiAlO{sub 2} as the tritium-producing material. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) is assumed as the calculation target, and using the continuous-energy Monte Carlo transport code MVP-BURN, burn-up simulations are carried out. To load sufficient Li into the core, LiAlO{sub 2} is loaded into the removable reflectors that surround the ring-shaped fuel blocks in addition to the burnable poison insertion holes. It is shown that module high-temperature gas-cooled reactors with a total thermal output power of 3 GW can produce almost 8 kg of tritium in a year.

  19. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  20. Gas-liquid mass transfer coefficient of methane in bubble column reactor

    International Nuclear Information System (INIS)

    Lee, Jaewon; Ha, Kyoung-Su; Lee, Jinwon; Kim, Choongik; Yasin, Muhammad; Park, Shinyoung; Chang, In Seop; Lee, Eun Yeol

    2015-01-01

    Biological conversion of methane gas has been attracting considerable recent interest. However, methanotropic bioreactor is limited by low solubility of methane gas in aqueous solution. Although a large mass transfer coefficient of methane in water could possibly overcome this limitation, no dissolved methane probe in aqueous environment is commercially available. We have developed a reactor enabling the measurement of aqueous phase methane concentration and mass transfer coefficient (k L a). The feasibility of the new reactor was demonstrated by measuring k L a values as a function of spinning rate of impeller and flow rate of methane gas. Especially, at spinning rate of 300 rpm and flow rate of 3.0 L/min, a large k L a value of 102.9 h -1 was obtained

  1. Estimating the composition of gas hydrate using 3D seismic data from Penghu Canyon, offshore Taiwan

    Directory of Open Access Journals (Sweden)

    Sourav Kumar Sahoo

    2018-01-01

    Full Text Available Direct measurements of gas composition by drilling at a few hundred meters below seafloor can be costly, and a remote sensing method may be preferable. The hydrate occurrence is seismically shown by a bottom-simulating reflection (BSR which is generally indicative of the base of the hydrate stability zone. With a good temperature profile from the seafloor to the depth of the BSR, a near-correct hydrate phase diagram can be calculated, which can be directly related to the hydrate composition. However, in the areas with high topographic anomalies of seafloor, the temperature profile is usually poorly defined, with scattered data. Here we used a remote method to reduce such scattering. We derived gas composition of hydrate in stability zone and reduced the scattering by considering depth-dependent geothermal conductivity and topographic corrections. Using 3D seismic data at the Penghu canyon, offshore SW Taiwan, we corrected for topographic focusing through 3D numerical thermal modeling. A temperature profile was fitted with a depth-dependent geothermal gradient, considering the increasing thermal conductivity with depth. Using a pore-water salinity of 2%, we constructed a gas hydrate phase model composed of 99% methane and 1% ethane to derive a temperature depth profile consistent with the seafloor temperature from in-situ measurements, and geochemical analyses of the pore fluids. The high methane content suggests predominantly biogenic source. The derived regional geothermal gradient is 40°C km-1. This method can be applied to other comparable marine environment to better constrain the composition of gas hydrate from BSR in a seismic data, in absence of direct sampling.

  2. Gas cooled reactor assessment. Volume II. Final report, February 9, 1976--June 30, 1976

    International Nuclear Information System (INIS)

    1976-08-01

    This report was prepared to document the estimated power plant capital and operating costs, and the safety and environmental assessments used in support of the Gas Cooled Reactor Assessment performed by Arthur D. Little, Inc. (ADL), for the U.S. Energy Research and Development Administration. The gas-cooled reactor technologies investigated include: the High Temperature Gas Reactor Steam Cycle (HTGR-SC), the HTGR Direct Cycle (HTGR-DC), the Very High Temperature Reactor (VHTR) and the Gas Cooled Fast Reactor (GCFR). Reference technologies used for comparison include: Light Water Reactors (LWR), the Liquid Metal Fast Breeder Reactor (LMFBR), conventional coal-fired steam plants, and coal combustion for process heat

  3. News from the world

    International Nuclear Information System (INIS)

    Anon.

    2006-01-01

    News from the world in relation with nuclear power and fuel cycle are given: Dismantling of the research reactor of the Pasteur Institute, Areva gets the contract to replace the vessel caps for the nuclear power plant of Diablo Canyon, the United Kingdom chooses the renewal of the nuclear park and an increase in the use of renewable energy sources, The united states launches a call to projects for the building of new generation nuclear power plants, in Argentina the government develops its nuclear industry, the Russian federation proposes the creation of an international center for the fuel cycle are the principal points that are developed in this issue. (N.C.)

  4. Calculation of gas-flow in plasma reactor for carbon partial oxidation

    Science.gov (United States)

    Bespala, Evgeny; Myshkin, Vyacheslav; Novoselov, Ivan; Pavliuk, Alexander; Makarevich, Semen; Bespala, Yuliya

    2018-03-01

    The paper discusses isotopic effects at carbon oxidation in low temperature non-equilibrium plasma at constant magnetic field. There is described routine of experiment and defined optimal parameters ensuring maximum enrichment factor at given electrophysical, gas-dynamic, and thermodymanical parameters. It has been demonstrated that at high-frequency generator capacity of 4 kW, supply frequency of 27 MHz and field density of 44 mT the concentration of paramagnetic heavy nuclei 13C in gaseous phase increases up to 1.78 % compared to 1.11 % for natural concentration. Authors explain isotopic effect decrease during plasmachemical separation induced by mixing gas flows enriched in different isotopes at the lack of product quench. With the help of modeling the motion of gas flows inside the plasma-chemical reactor based on numerical calculation of Navier-Stokes equation authors determine zones of gas mixing and cooling speed. To increase isotopic effects and proportion of 13C in gaseous phase it has been proposed to use quench in the form of Laval nozzle of refractory steel. The article represents results on calculation of optimal Laval Nozzle parameters for plasma-chemical reactor of chosen geometry of. There are also given dependences of quench time of products on pressure at the diffuser output and on critical section diameter. Authors determine the location of quench inside the plasma-chemical reactor in the paper.

  5. Canyons off northwest Puerto Rico

    International Nuclear Information System (INIS)

    Gardner, W.D.; Glover, L.K.; Hollister, C.D.

    1980-01-01

    The Nuclear-Research Submarine NR-1 was used to study morphoplogy, sediment, and sediment-water interactions off the northwest coast of Puerto Rico. New detailed bathymetry from the surface-support ship, USS Portland, shows several submarine canyons in the area, some of them unreported previously. The north coast canyons, Arecibo, Tiberones and Quebradillas, are primarily erosional features although no recent turbidity-current evidence is seen. The canyons are presently filling with river-transported sediments. (orig./ME)

  6. Rotating bed reactor for CLC: Bed characteristics dependencies on internal gas mixing

    International Nuclear Information System (INIS)

    Håkonsen, Silje Fosse; Grande, Carlos A.; Blom, Richard

    2014-01-01

    Highlights: • A mathematical model for the rotating CLC reactor has been developed. • The model reflects the gas distribution in the reactor during CLC operation. • Radial dispersion in the rotating bed is the main cause for internal gas mixing. • The model can be used to optimize the reactor design and particle characteristics. - Abstract: A newly designed continuous lab-scale rotating bed reactor for chemical looping combustion using CuO/Al 2 O 3 oxygen carrier spheres and methane as fuel gives around 90% CH 4 conversion and >90% CO 2 capture efficiency based on converted methane at 800 °C. However, from a series of experiments using a broad range of operating conditions potential CO 2 purities only in the range 20–65% were yielded, mostly due to nitrogen slip from the air side of the reactor into the effluent CO 2 stream. A mathematical model was developed intending to understand the air-mixing phenomena. The model clearly reflects the gas slippage tendencies observed when varying the process conditions such as rotation frequency, gas flow and the flow if inert gas in the two sectors dividing the air and fuel side of the reactor. Based on the results, it is believed that significant improvements can be made to reduce gas mixing in future modified and scaled-up reactor versions

  7. Gas Nozzle Effect on the Deposition of Polysilicon by Monosilane Siemens Reactor

    Directory of Open Access Journals (Sweden)

    Seung Oh Kang

    2012-01-01

    Full Text Available Deposition of polysilicon (poly-Si was tried to increase productivity of poly-Si by using two different types of gas nozzle in a monosilane Bell-jar Siemens (MS-Siemens reactor. In a mass production of poly-Si, deposition rate and energy consumption are very important factors because they are main performance indicators of Siemens reactor and they are directly related with the production cost of poly-Si. Type A and B nozzles were used for investigating gas nozzle effect on the deposition of poly-Si in a MS-Siemens reactor. Nozzle design was analyzed by computation cluid dynamics (CFD. Deposition rate and energy consumption of poly-Si were increased when the type B nozzle was used. The highest deposition rate was 1 mm/h, and the lowest energy consumption was 72 kWh⋅kg-1 in this study.

  8. Effects of trees on the dilution of vehicle exhaust emissions in urban street canyons

    NARCIS (Netherlands)

    Gromke, C.B.; Ruck, B.

    2009-01-01

    In order to investigate the natural ventilation and air quality of urban street canyons with trees, boundary layer wind tunnel studies at a small-scale model have been performed. Concentrations in street canyons with a tracer gas emitting line source at the ground level and one row of trees arranged

  9. Analysis of Neutron Flux Distribution in Rsg-Gas Reactor With U-Mo Fuels

    Directory of Open Access Journals (Sweden)

    Taswanda Taryo

    2004-01-01

    Full Text Available The use of U-Mo fuels in research reactors seems to be promising and, recently, world researchers have carried out these such activities actively. The National Nuclear Energy Agency (BATAN which owns RSG-GAS reactor available in Serpong Research Center for Atomic Energy should anticipate this trend. It is, therefore, this research work on the use of U-Mo fuels in RSG-GAS reactor should be carried out. The work was focused on the analysis of neutron flux distribution in the RSG-GAS reactor using different content of molybdenum in U-Mo fuels. To begin with, RSG-GAS reactor core model was developed and simulated into X, Y and Z dimensions. Cross section of materials based on the developed cells of standard and control fuels was then generated using WIMS-D5-B. The criticality calculations were finally carried out applying BATAN-2DIFF code. The results showed that the neutron flux distribution obtained in U-Mo-fuel-based RSG-GAS core is very similar to those achieved in the 300-gram sillicide-fuel-based RSG-GAS reactor core. Indeed, the utilization of the U-Mo RSG-GAS core can be very similar to that of the high-density sillicide reactor core and even could be better in the future.

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

    International Nuclear Information System (INIS)

    2009-01-01

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

  11. Summary of ORNL high-temperature gas-cooled reactor program

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

    Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

  12. Fast reactor cover gas purification - The UK position

    Energy Technology Data Exchange (ETDEWEB)

    Thorley, A W

    1987-07-01

    The cover gas in the Prototype Fast Reactor (PFR) provides an inert gas blanket for both primary and secondary sodium circuits, ensures inert gas padding exists between the upper seals associated with penetrations through the reactor roof and provides argon to items of plant such as the control rods and the rotating shield and also to on line instruments such as the secondary circuit Katharometers. In order to meet these and other requirements purification of the argon cover gas is important to ensure: gas fed to purge gaps in the area of the magnetic hold device in the control rod mechanisms is not laden with sodium aerosols and reactive impurities (O{sub 2}, H{sub 2}) which could cause blocking both within the gaps and pipelines; gas phase detection systems which provide early warning of steam generator failures or oil ingress into the sodium are not affected by the presence of gaseous impurities such as H{sub 2}, CO/CO{sub 2} and CH{sub 4}; mass transfer processes involving both corrosion products and interstitial atoms cannot be sustained in the cover gas environment due to the presence of high levels of O{sub 2}, N{sub 2} and carburising gases; background levels of radioactivity (eg Xe 133) are sufficiently low to enable gas phase detection of failed fuel pins, and the primary circuit gas blanket activity is sufficiently reduced so that discharges to the atmosphere are minimised. This paper describes how the PFR cover gas purification system is coping with these various items and how current thinking regarding the design of cover gas purification systems for a Civil Demonstration Fast Reactor (CDFR), where larger gas volumes and higher levels of radioactivity may be involved, is being guided by current experience on PFR. The paper also briefly review the experimental work planned to study aerosol and caesium behaviour in cove gas environments and discusses the behaviour of those impurities such as Zn, oil and N{sub 2} which are potentially damaging if certain

  13. Fast reactor cover gas purification - The UK position

    International Nuclear Information System (INIS)

    Thorley, A.W.

    1987-01-01

    The cover gas in the Prototype Fast Reactor (PFR) provides an inert gas blanket for both primary and secondary sodium circuits, ensures inert gas padding exists between the upper seals associated with penetrations through the reactor roof and provides argon to items of plant such as the control rods and the rotating shield and also to on line instruments such as the secondary circuit Katharometers. In order to meet these and other requirements purification of the argon cover gas is important to ensure: gas fed to purge gaps in the area of the magnetic hold device in the control rod mechanisms is not laden with sodium aerosols and reactive impurities (O 2 , H 2 ) which could cause blocking both within the gaps and pipelines; gas phase detection systems which provide early warning of steam generator failures or oil ingress into the sodium are not affected by the presence of gaseous impurities such as H 2 , CO/CO 2 and CH 4 ; mass transfer processes involving both corrosion products and interstitial atoms cannot be sustained in the cover gas environment due to the presence of high levels of O 2 , N 2 and carburising gases; background levels of radioactivity (eg Xe 133) are sufficiently low to enable gas phase detection of failed fuel pins, and the primary circuit gas blanket activity is sufficiently reduced so that discharges to the atmosphere are minimised. This paper describes how the PFR cover gas purification system is coping with these various items and how current thinking regarding the design of cover gas purification systems for a Civil Demonstration Fast Reactor (CDFR), where larger gas volumes and higher levels of radioactivity may be involved, is being guided by current experience on PFR. The paper also briefly review the experimental work planned to study aerosol and caesium behaviour in cove gas environments and discusses the behaviour of those impurities such as Zn, oil and N 2 which are potentially damaging if certain levels are exceeded in operating

  14. Modeling and performance of the MHTGR [Modular High-Temperature Gas-Cooled Reactor] reactor cavity cooling system

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1990-04-01

    The Reactor Cavity Cooling System (RCCS) of the Modular High- Temperature Gas-Cooled Reactor (MHTGR) proposed by the U.S. Department of Energy is designed to remove the nuclear afterheat passively in the event that neither the heat transport system nor the shutdown cooling circulator subsystem is available. A computer dynamic simulation for the physical and mathematical modeling of and RCCS is described here. Two conclusions can be made form computations performed under the assumption of a uniform reactor vessel temperature. First, the heat transferred across the annulus from the reactor vessel and then to ambient conditions is very dependent on the surface emissivities of the reactor vessel and RCCS panels. These emissivities should be periodically checked to ensure the safety function of the RCCS. Second, the heat transfer from the reactor vessel is reduced by a maximum of 10% by the presence of steam at 1 atm in the reactor cavity annulus for an assumed constant in the transmission of radiant energy across the annulus can be expected to result in an increase in the reactor vessel temperature for the MHTGR. Further investigation of participating radiation media, including small particles, in the reactor cavity annulus is warranted. 26 refs., 7 figs., 1 tab

  15. Study on thermodynamic cycle of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu Xinhe; Yang Xiaoyong; Wang Jie

    2017-01-01

    The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

  16. Gas-liquid mass transfer coefficient of methane in bubble column reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jaewon; Ha, Kyoung-Su; Lee, Jinwon; Kim, Choongik [Sogang University, Seoul (Korea, Republic of); Yasin, Muhammad; Park, Shinyoung; Chang, In Seop [Gwangju Institute of Science and Technology (GIST), Gwangju (Korea, Republic of); Lee, Eun Yeol [Kyung Hee University, Yongin (Korea, Republic of)

    2015-06-15

    Biological conversion of methane gas has been attracting considerable recent interest. However, methanotropic bioreactor is limited by low solubility of methane gas in aqueous solution. Although a large mass transfer coefficient of methane in water could possibly overcome this limitation, no dissolved methane probe in aqueous environment is commercially available. We have developed a reactor enabling the measurement of aqueous phase methane concentration and mass transfer coefficient (k{sub L}a). The feasibility of the new reactor was demonstrated by measuring k{sub L}a values as a function of spinning rate of impeller and flow rate of methane gas. Especially, at spinning rate of 300 rpm and flow rate of 3.0 L/min, a large k{sub L}a value of 102.9 h{sup -1} was obtained.

  17. Insight conference proceedings : Pacific Canada and north coast offshore oil and gas development

    International Nuclear Information System (INIS)

    2005-01-01

    This conference provided a forum for reviewing oil and gas activities in offshore basins around the world and their influence on British Columbia's plans to reopen its offshore to exploration and development. Details on jurisdictional issues were provided along with potential environmental impacts and remedial actions. Speakers with first hand experience in the latest technologies and mitigation measures addressed issues regarding regulatory regimes used in the Arctic offshore and compared them with proposed developments in British Columbia. Issues specific to the Pacific coast were also discussed, such as the current offshore oil and gas moratorium, land use and the views of First Nations and environmental non-governmental organizations. A current industry perspective of offshore oil and gas development was presented with lessons learned from the east coast. The opportunities for constructing liquefied natural gas (LNG) terminals and their associated facilities in the Pacific coast were also presented. This conference featured 14 presentations, of which 3 have been indexed separately for inclusion in this database

  18. A novel approach for toluene gas treatment using a downflow hanging sponge reactor.

    Science.gov (United States)

    Yamaguchi, Tsuyoshi; Nakamura, Syoichiro; Hatamoto, Masashi; Tamura, Eisuke; Tanikawa, Daisuke; Kawakami, Shuji; Nakamura, Akinobu; Kato, Kaoru; Nagano, Akihiro; Yamaguchi, Takashi

    2018-05-01

    A novel gas-scrubbing bioreactor based on a downflow hanging sponge (DHS) reactor was developed as a new volatile organic compound (VOC) treatment system. In this study, the effects of varying the space velocity and gas/liquid ratio were investigated to assess the effectiveness of using toluene gas as a model VOC. Under optimal conditions, the toluene removal rate was greater than 80%, and the maximum elimination capacity was observed at approximately 13 g-C m -3  h -1 . The DHS reactor demonstrated slight pressure loss (20 Pa) and a high concentration of suspended solids (up to 30,000 mg/L-sponge). Cloning analysis of the 16S rRNA and functional genes of toluene degradation pathways (tmoA, todC, tbmD, xylA, and bssA) revealed that the clones belonging to the toluene-degrading bacterium Pseudomonas putida constituted the predominant species detected at the bottom of the DHS reactor. The toluene-degrading bacteria Pseudoxanthomonas spadix and Pseudomonas sp. were also detected by tmoA- and todC-targeted cloning analyses, respectively. These results demonstrate the potential for the industrial application of this novel DHS reactor for toluene gas treatment.

  19. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    Markoczy, G.; Hudina, M.; Richmond, R.; Wydler, P.; Stratton, R.W.; Burgsmueller, P.

    1980-03-01

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1979 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  20. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sandalls, F.J.

    1978-03-01

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10 -9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  1. The variation of particle gas-borne concentration with time in a gas cooled reactor

    International Nuclear Information System (INIS)

    Reed, J.; Hall, D.; Reeks, M.W.

    1985-01-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  2. The variation of particle gas-borne concentration with time in a gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Reed, J; Hall, D; Reeks, M W [Central Electricity Generating Board, Berkeley Nuclear Laboratories (United Kingdom)

    1985-07-01

    If volatile fission products are released from fuel during a reactor fault, a significant fraction could become attached to small particles also present in the coolant. In such circumstances the retention of those particles by the reactor circuit will limit the level of gas-borne particle concentration and hence be important in reducing the potential release of fission product activity to the atmosphere. Clearly the retention of particles will be influenced by both the deposition and resuspension of particles from surfaces exposed to the coolant flow. In this paper we consider deposition and resuspension but pay particular attention to the role of resuspension, which in the past has been given little consideration. A recently developed model for the resuspension of small particles by a turbulent flow is outlined. Traditionally, resuspension has been interpreted as a force balance between the aerodynamic removal forces and the surface adhesive forces. In contrast, this new approach embodies an energy balance criterion for particle resuspension. Furthermore, the stochastic nature of this new model has shown that resuspension can be sub-divided into two regimes: (i) initial resuspension (resuspension occurring in times less than a second) which reduces the net deposition of particles to a surface; and (ii) longer term resuspension (resuspension after 1 second) which determines the asymptotic decay of particle gas-borne concentration. It is seen that the asymptotic decay varies almost inversely as the decay time. Force balance models are unsuccessful in accounting for the experimentally observed longer term resuspension. We show that a Volterra integro-differential equation best describes the variation of particle gas-borne concentration with time in a recirculating gas flow such as a gas cooled reactor. It is seen that the longer term resuspension has a major influence in the final decay of particle concentration. (author)

  3. Design activity of IHI on the experimental multipurpose high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    1978-01-01

    With conspicuous interest and attention paid by iron and steel manufacturing industries, the development of the multipurpose high temperature gas-cooled reactor, namely the process heat reactor has been energetically discussed in Japan. The experimental multipurpose high temperature gas-cooled reactor, planned by JAERI (the Japan Atomic Energy Research Institute), is now at the end of the adjustment design stage and about to enter the system synthesizing design stage. The design of the JAERI reactor as a pilot plant for process heat reactors that make possible the direct use of the heat, produced in the reactor, for other industrial uses was started in 1969, and has undergone several revisions up to now. The criticality of the JAERI reactor is expected to be realized before 1985 according to the presently published program. IHI has engaged in the developing work of HTGR (high temperature gas-cooled reactor) including VHTR (very high temperature gas-cooled reactor) for over seven years, producing several achievements. IHI has also participated in the JAERI project since 1973 with some other companies concerned in this field. The design activity of IHI in the development of the JAERI reactor is briefly presented in this paper. (auth.)

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

    International Nuclear Information System (INIS)

    Surian P, Tukiran S

    1999-01-01

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

  5. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  6. Structure-forming corals and sponges and their use as fish habitat in Bering Sea submarine canyons.

    Science.gov (United States)

    Miller, Robert J; Hocevar, John; Stone, Robert P; Fedorov, Dmitry V

    2012-01-01

    Continental margins are dynamic, heterogeneous settings that can include canyons, seamounts, and banks. Two of the largest canyons in the world, Zhemchug and Pribilof, cut into the edge of the continental shelf in the southeastern Bering Sea. Here currents and upwelling interact to produce a highly productive area, termed the Green Belt, that supports an abundance of fishes and squids as well as birds and marine mammals. We show that in some areas the floor of these canyons harbors high densities of gorgonian and pennatulacean corals and sponges, likely due to enhanced surface productivity, benthic currents and seafloor topography. Rockfishes, including the commercially important Pacific ocean perch, Sebastes alutus, were associated with corals and sponges as well as with isolated boulders. Sculpins, poachers and pleuronectid flounders were also associated with corals in Pribilof Canyon, where corals were most abundant. Fishes likely use corals and sponges as sources of vertical relief, which may harbor prey as well as provide shelter from predators. Boulders may be equivalent habitat in this regard, but are sparse in the canyons, strongly suggesting that biogenic structure is important fish habitat. Evidence of disturbance to the benthos from fishing activities was observed in these remote canyons. Bottom trawling and other benthic fishing gear has been shown to damage corals and sponges that may be very slow to recover from such disturbance. Regulation of these destructive practices is key to conservation of benthic habitats in these canyons and the ecosystem services they provide.

  7. H Canyon Processing In Correlation With FH Analytical Labs

    International Nuclear Information System (INIS)

    Weinheimer, E.

    2012-01-01

    Management of radioactive chemical waste can be a complicated business. H Canyon and F/H Analytical Labs are two facilities present at the Savannah River Site in Aiken, SC that are at the forefront. In fact H Canyon is the only large-scale radiochemical processing facility in the United States and this processing is only enhanced by the aid given from F/H Analytical Labs. As H Canyon processes incoming materials, F/H Labs provide support through a variety of chemical analyses. Necessary checks of the chemical makeup, processing, and accountability of the samples taken from H Canyon process tanks are performed at the labs along with further checks on waste leaving the canyon after processing. Used nuclear material taken in by the canyon is actually not waste. Only a small portion of the radioactive material itself is actually consumed in nuclear reactors. As a result various radioactive elements such as Uranium, Plutonium and Neptunium are commonly found in waste and may be useful to recover. Specific processing is needed to allow for separation of these products from the waste. This is H Canyon's specialty. Furthermore, H Canyon has the capacity to initiate the process for weapons-grade nuclear material to be converted into nuclear fuel. This is one of the main campaigns being set up for the fall of 2012. Once usable material is separated and purified of impurities such as fission products, it can be converted to an oxide and ultimately turned into commercial fuel. The processing of weapons-grade material for commercial fuel is important in the necessary disposition of plutonium. Another processing campaign to start in the fall in H Canyon involves the reprocessing of used nuclear fuel for disposal in improved containment units. The importance of this campaign involves the proper disposal of nuclear waste in order to ensure the safety and well-being of future generations and the environment. As processing proceeds in the fall, H Canyon will have a substantial

  8. Fast reactor primary cover gas system proposals for CDFR

    International Nuclear Information System (INIS)

    Harrison, L.M.T.

    1987-01-01

    A primary sodium gas cover has been designed for CDFR, it comprises plant to maintain and control; cover gas pressure for all reactor operating at fault conditions, cover gas purity by both blowdown and by a special clean-up facility and the clean argon supply for the failed fuel detection system and the primary pump seal purge. The design philosophy is to devise a cover gas system that can be specified for any LMFBR where only features like vessel and pipework size need to be altered to suit different design and operating conditions. The choice of full power and shutdown operating pressures is derived and the method chosen to control these values is described. A part active/part passive system is proposed for this duty, a surge volume of 250 m 3 gives passive control between full power and hot shutdown. Pressure control operation criteria is presented for various reactor operating conditions. A design for a sodium aerosol filter, based on that used on PFR is presented, it is specifically designed so that it can be fitted with an etched disc type particulate filter and maintenance is minimised. Two methods that maintain cover gas purity are described. The first, used during normal reactor operation with a small impurities ingress, utilises the continuous blowdown associated with the inevitable clean argon purge through the various reactor component seals. The second method physically removes the impurities xenon and krypton from the cover gas by their adsorption, at cryogenic temperature, onto a bed of activated carbon. The equipment required for these two duties and their mode of operation is described with the aid of a system flow diagram. The primary pump seals requires a gas purge to suppress aerosol migration. A system where the argon used for this task is recirculated and partially purified is described. (author)

  9. Formative flow in bedrock canyons

    Science.gov (United States)

    Venditti, J. G.; Kwoll, E.; Rennie, C. D.; Church, M. A.

    2017-12-01

    In alluvial channels, it is widely accepted that river channel configuration is set by a formative flow that represents a balance between the magnitude and frequency of flood flows. The formative flow is often considered to be one that is just capable of filling a river channel to the top of its banks. Flows much above this formative flow are thought to cause substantial sediment transport and rearrange the channel morphology to accommodate the larger flow. This idea has recently been extended to semi-alluvial channels where it has been shown that even with bedrock exposed, the flows rarely exceed that required to entrain the local sediment cover. What constitutes a formative flow in a bedrock canyon is not clear. By definition, canyons have rock walls and are typically incised vertically, removing the possibility of the walls being overtopped, as can occur in an alluvial channel at high flows. Canyons are laterally constrained, have deep scour pools and often have width to maximum depth ratios approaching 1, an order of magnitude lower than alluvial channels. In many canyons, there are a sequence of irregularly spaced scour pools. The bed may have intermittent or seasonal sediment cover, but during flood flows the sediment bed is entrained leaving a bare bedrock channel. It has been suggested that canyons cut into weak, well-jointed rock may adjust their morphology to the threshold for block plucking because the rock bed is labile during exceptionally large magnitude flows. However, this hypothesis does not apply to canyons cut into massive crystalline rock where abrasion is the dominant erosion process. Here, we argue that bedrock canyon morphology is adjusted to a characteristic flow structure developed in bedrock canyons. We show that the deeply scoured canyon floor is adjusted to a velocity inversion that is present at low flows, but gets stronger at high flows. The effect is to increase boundary shear stresses along the scour pool that forms in constricted

  10. Principle and Performance of Gas Self-inducing Reactors and Applications to Biotechnology.

    Science.gov (United States)

    Ye, Qin; Li, Zhimin; Wu, Hui

    2016-01-01

    Gas-liquid contacting is an important unit operation in chemical and biochemical processes, but the gas utilization efficiency is low in conventional gas-liquid contactors especially for sparingly soluble gases. The gas self-inducing impeller is able to recycle gas in the headspace of a reactor to the liquid without utilization of additional equipment such as a gas compressor, and thus, the gas utilization efficiency is significantly enhanced. Gas induction is caused by the low pressure or deep vortex at a sufficiently high impeller speed, and the speed at which gas induction starts is termed the critical speed. The critical impeller speed, gas-induction flow rate, power consumption, and gas-liquid mass transfer are determined by the impeller design and operation conditions. When the reactor is operated in a dead-end mode, all the introduced gas can be completely used, and this feature is especially favorable to flammable and/or toxic gases. In this article, the principles, designs, characteristics of self-inducing reactors, and applications to biotechnology are described.

  11. Plasmachemical Oxidation Processes in Hybrid Gas-Liquid Electrical Discharge Reactor

    Czech Academy of Sciences Publication Activity Database

    Lukeš, Petr; Locke, B.R.

    2005-01-01

    Roč. 38, č. 22 (2005), s. 4074-4081 ISSN 0022-3727 Institutional research plan: CEZ:AV0Z20430508 Keywords : Corona discharge * hybrid reactor * hydroxyl radical * ozone * phenol * water treatment Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.957, year: 2005

  12. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1982-01-01

    Towards the second half of a fifty-year time span the market for gas-cooled reactors as sources of high temperature process heat and as highly fuel efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realisation of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel-making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  13. Long-term prospects for the gas-cooled reactor

    International Nuclear Information System (INIS)

    Tan, W.P.S.

    1983-01-01

    Towards the second half of a 50-year time span the market for gas-cooled reactors as sources of high-temperature process heat and as highly fuel-efficient electricity producers should be reasonably bright, given a fair degree of technological maturity and consequent realization of inherent economic advantages. Declining fossil resources and increasing prices, initially in oil and gas, later in open-cast coal, provide the economic impetus towards substitution of nuclear for coal heat, not only in the generally accepted processes of coal conversion and steel making but also for oil shale pyrolysis and electrothermal aluminium smelting. Around 2010, if not sooner, the need for uranium conservation should allow the market penetration of breeders and thorium-cycle reactors for which gas cooling has a potential techno-economic edge. (author)

  14. 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.

  15. Design and development of gas turbine high temperature reactor 300 (GTHTR300)

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Katanishi, Shoji; Takada, Shoji; Takizuka, Takakazu; Yan, Xing; Kosugiyama, Shinichi

    2003-01-01

    JAERI (Japan Atomic Energy Research Institute) started design and development of the high temperature gas cooled reactor with a gas turbine electric generation system, GTHTR300, in April 2001. Design originalities of the GTHTR300 are a horizontally mounted highly efficient gas turbine system and an ultimately simplified safety system such as no containment building and no active emergency core cooling. These design originalities are proposed based on design and operational experiences in conventional gas turbine systems and Japan's first high temperature gas cooled reactor (HTTR: High Temperature Engineering Test Reactor) so that many R and Ds are not required for the development. Except these original design features, devised core design, fuel design and plant design are adopted to meet design requirements and attain a target cost. This paper describes the unique design features focusing on the safety design, reactor core design and gas turbine system design together with a preliminary result of the safety evaluation carried out for a typical severe event. This study is entrusted from Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

  16. The Gas Turbine - Modular Helium Reactor: A Promising Option for Near Term Deployment

    International Nuclear Information System (INIS)

    LaBar, Malcolm P.

    2002-01-01

    The Gas Turbine - Modular Helium Reactor (GT-MHR) is an advanced nuclear power system that offers unparalleled safety, high thermal efficiency, environmental advantages, and competitive electricity generation costs. The GT-MHR module couples a gas-cooled modular helium reactor (MHR) with a high efficiency modular Brayton cycle gas turbine (GT) energy conversion system. The reactor and power conversion systems are located in a below grade concrete silo that provides protection against sabotage. The GT-MHR safety is achieved through a combination of inherent safety characteristics and design selections that take maximum advantage of the gas-cooled reactor coated particle fuel, helium coolant and graphite moderator. The GT-MHR is projected to be economically competitive with alternative electricity generation technologies due to the high operating temperature of the gas-cooled reactor, high thermal efficiency of the Brayton cycle power conversion system, high fuel burnup (>100,000 MWd/MT), and low operation and maintenance requirements. (author)

  17. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Bjornard, Trond; Hockert, John

    2011-01-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC and A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC and A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC and A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR (Pty) and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC and A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR and D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present

  18. Method of collecting helium cover gas for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Miyamoto, Keiji; Ueda, Hiroshi.

    1981-01-01

    Purpose: To reduce the systematic facility cost in a heavy water moderated reactor by contriving the simplification of a helium cover gas collecting intake system. Method: A detachable low pressure metal tank and a neoprene balloon are prepared for a vacuum pump in a permanent vacuum drying facility. When all of the helium cover gas is collected from a heavy water moderated reactor, a large capacity of neoprene balloon capable of temporarily storing it under low pressure is connected to the exhaust of the vacuum pump. On the other hand, while the reactor is operating, a suitable amount of the low pressure tank or neoprene balloon is connected to the exhaust side of the pump, thereby regulating the pressure of the helium cover gas. When refeeding the cover gas, the balloon, with a large capacity for collecting and storing the cover gas is connected to the intake side of the pump. Thus, the pressure regulation, collection of all of the cover gas and refeeding of the cover gas can be conducted without using a high discharge pump and high pressure tank. (Kamimura, M.)

  19. Thermodynamic properties of helium in the range from 20 to 15000C and 1 to 100 bar. Reactor core design of high-temperature gas-cooled reactors. Pt. 1

    International Nuclear Information System (INIS)

    Kipke, H.E.; Stoehr, A.; Banerjea, A.; Hammeke, K.; Huepping, N.

    1978-12-01

    The following report presents in tabular form the safety standard of the nuclear safety standard commission (KTA) on reactor core design of high-temperature gas-cooled reactors. Part 1: Calculation of thermodynamic properties of helium The basis of the present work is the data and formulae given by H. Petersen for the calculation of density, specific heat, thermal conductivity and dynamic viscosity of helium together with the formula for their standard deviations in the range of temperature and pressure stated above. The relations for specific enthalpy and specific entropy have been derived from density and specific heat, whereby specific heat is assumed constant over the given range of temperature and pressure. The latter section of this report contains tables of thermodynamic properties of helium calculated from the equations stated earlier in this paper. (orig.) [de

  20. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  1. Hydrogen production with fully integrated fuel cycle gas and vapour core reactors

    International Nuclear Information System (INIS)

    Anghaie, S.; Smith, B.

    2004-01-01

    This paper presents results of a conceptual design study involving gas and vapour core reactors (G/VCR) with a combined scheme to generate hydrogen and power. The hydrogen production schemes include high temperature electrolysis as well as two dominant thermochemical hydrogen production processes. Thermochemical hydrogen production processes considered in this study included the calcium-bromine process and the sulphur-iodine processes. G/VCR systems are externally reflected and moderated nuclear energy systems fuelled by stable uranium compounds in gaseous or vapour phase that are usually operated at temperatures above 1500 K. A gas core reactor with a condensable fuel such as uranium tetrafluoride (UF 4 ) or a mixture of UF 4 and other metallic fluorides (BeF 2 , LiF, KF, etc.) is commonly known as a vapour core reactor (VCR). The single most relevant and unique feature of gas/vapour core reactors is that the functions of fuel and coolant are combined into one. The reactor outlet temperature is not constrained by solid fuel-cladding temperature limits. The maximum fuel/working fluid temperature in G/VCR is only constrained by the reactor vessel material limits, which is far less restrictive than the fuel clad. Therefore, G/VCRs can potentially provide the highest reactor and cycle temperature among all existing or proposed fission reactor designs. Gas and vapour fuel reactors feature very low fuel inventory and fully integrated fuel cycle that provide for exceptional sustainability and safety characteristics. With respect to fuel utilisation, there is no fuel burn-up limit for gas core reactors due to continuous recycling of the fuel. Owing to the flexibility in nuclear design characteristics of cavity reactors, a wide range of conversion ratio from completely burner to breeder is achievable. The continuous recycling of fuel in G/VCR systems allow for complete burning of actinides without removing and reprocessing of the fuel. The only waste products at the back

  2. Geologic quadrangle maps of the United States: geology of the Casa Diablo Mountain quadrangle, California

    Science.gov (United States)

    Rinehart, C. Dean; Ross, Donald Clarence

    1957-01-01

    The Casa Diablo Mountain quadrangle was mapped in the summers of 1952 and 1953 by the U.S. Geological Survey in cooperation with the California State Division of Mines as part of a study of potential tungsten-bearing areas.

  3. Operating US power reactors

    International Nuclear Information System (INIS)

    Silver, E.G.

    1988-01-01

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

  4. Hydrodynamic flow regimes, gas holdup, and liquid circulation in airlift reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abashar, M.E.; Narsingh, U.; Rouillard, A.E.; Judd, R. [Univ. of Durban (South Africa)

    1998-04-01

    This study reports an experimental investigation into the hydrodynamic behavior of an external-loop airlift reactor (ALR) for the air-water system. Three distinct flow regimes are identified--namely homogeneous, transition, and heterogeneous regimes. The transition between homogeneous and heterogeneous flow is observed to occur over a wide range rather than being merely a single point as has been previously reported in the literature. A gas holdup correlation is developed for each flow regime. The correlations fit the experimental gas holdup data with very good accuracy (within {+-}5%). It would appear, therefore, that a deterministic equation to describe each flow regime is likely to exist in ALRs. This equation is a function of the reactor geometry and the system`s physical properties. New data concerning the axial variation of gas holdup is reported in which a minimum value is observed. This phenomenon is discussed and an explanation offered. Discrimination between two sound theoretical models--namely model 1 (Chisti et al., 1988) and model 2 (Garcia Calvo, 1989)--shows that model 1 predicts satisfactorily the liquid circulation velocity with an error of less than {+-} 10%. The good predictive features of model 1 may be due to the fact that it allows for a significant energy dissipation by wakes behind bubbles. Model 1 is now further improved by the new gas holdup correlations which are derived for the three different flow regimes.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-07-01

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

  6. The Black Canyon of the Gunnison: Today and Yesterday

    Science.gov (United States)

    Hansen, Wallace R.

    1965-01-01

    Since the early visit of Captain John William Gunnison in the middle of the last century, the Black Canyon of the Gunnison has stirred mixed apprehension and wonder in the hearts of its viewers. It ranks high among the more awesome gorges of North America. Many great western canyons are as well remembered for their brightly colored walls as for their airy depths. Not so the Black Canyon. Though it is assuredly not black, the dark-gray tones of its walls and the hazy shadows of its gloomy depths join together to make its name well deserved. Its name conveys an impression, not a picture. After the first emotional impact of the canyon, the same questions come to the minds of most reflective viewers and in about the following order: How deep is the Black Canyon, how wide, how does it compare with other canyons, what are the rocks, how did it form, and how long did it take? Several western canyons exceed the Black Canyon in overall size. Some are longer; some are deeper; some are narrower; and a few have walls as steep. But no other canyon in North American combines the depth, narrowness, sheerness, and somber countenance of the Black Canyon. In many places the Black Canyon is as deep as it is wide. Between The Narrows and Chasm View in the Black Canyon of the Gunnison National Monument (fig. 15) it is much deeper than wide. Average depth in the monument is about 2,000 feet, ranging from a maximum of about 2,700 feet, north of Warner Point (which also is the greatest depth anywhere in the canyon), to a minimum of about 1,750 feet at The Narrows. The stretch of canyon between Pulpit Rock and Chasm View, including The Narrows, though the shallowest in the monument, is also the narrowest, has some of the steepest walls, and is, therefore, among the most impressive segments of the canyon (fig. 3). Profiles of several well-known western canyons are shown in figure 1. Deepest of these by far is Hells Canyon of the Snake, on the Idaho-Oregon border. Clearly, it dwarfs the

  7. Canyon air flow measurement utilizing ASME standard pitot tube arrays

    International Nuclear Information System (INIS)

    Moncrief, B.R.

    1990-01-01

    The Savannah River Site produces nuclear materials for national defense. In addition to nuclear reactors, the site has separation facilities for reprocessing irradiated nuclear fuel. The chemical separation of highly radioactive materials takes place by remote control in large buildings called canyons. Personnel in these buildings are shielded from radiation by thick concrete walls. Contaminated air is exhausted from the canyons and contaminants are removed by sand filters prior to release to the atmosphere through a stack. When these facilities were built on a crash basis in the early 1950's, inadequate means were provided for pressure and air flow measurement. This presentation describes the challenge we faced in retrofitting a highly radioactive, heavily shielded facility with instrumentation to provide this capability

  8. Structure-forming corals and sponges and their use as fish habitat in Bering Sea submarine canyons.

    Directory of Open Access Journals (Sweden)

    Robert J Miller

    Full Text Available Continental margins are dynamic, heterogeneous settings that can include canyons, seamounts, and banks. Two of the largest canyons in the world, Zhemchug and Pribilof, cut into the edge of the continental shelf in the southeastern Bering Sea. Here currents and upwelling interact to produce a highly productive area, termed the Green Belt, that supports an abundance of fishes and squids as well as birds and marine mammals. We show that in some areas the floor of these canyons harbors high densities of gorgonian and pennatulacean corals and sponges, likely due to enhanced surface productivity, benthic currents and seafloor topography. Rockfishes, including the commercially important Pacific ocean perch, Sebastes alutus, were associated with corals and sponges as well as with isolated boulders. Sculpins, poachers and pleuronectid flounders were also associated with corals in Pribilof Canyon, where corals were most abundant. Fishes likely use corals and sponges as sources of vertical relief, which may harbor prey as well as provide shelter from predators. Boulders may be equivalent habitat in this regard, but are sparse in the canyons, strongly suggesting that biogenic structure is important fish habitat. Evidence of disturbance to the benthos from fishing activities was observed in these remote canyons. Bottom trawling and other benthic fishing gear has been shown to damage corals and sponges that may be very slow to recover from such disturbance. Regulation of these destructive practices is key to conservation of benthic habitats in these canyons and the ecosystem services they provide.

  9. State of development of gas cooled reactors in the Union of Soviet Socialist Republics

    International Nuclear Information System (INIS)

    Grebennik, V.N.; Mosevitskij, I.S.

    1991-01-01

    In the context of the programme for the development of gas-cooled reactors in the USSR it is reported that pilot plants with VGR-50 MW(el) and VG-400 MW(el) have been developed up to the stage of engineering design and that now the efforts are concentrated on the project of pilot-commercial reactor plant VGM (PCRP VGM) of a modular type with unit thermal power of 200-250 MW. The installation is designed to solve the main scientific and engineering problems of construction of high-temperature gas-cooled reactors, to test equipment components, and to show advantages of the given type of installations having the enhanced safety and capability to generate high-potential heat. The status of work on the PCRP VGM project is described. 3 refs, 1 fig., 1 tab

  10. Analysis of Kinetic Parameter Effect on Reactor Operation Stability of the RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Rokhmadi

    2007-01-01

    Kinetic parameter has influence to behaviour on RSG-GAS reactor operation. In this paper done is the calculation of reactivity curve, period-reactivity relation and low power transfer function in silicide fuel. This parameters is necessary and useful for reactivity characteristic analysis and reactor stability. To know the reactivity response, it was done reactivity insertion at power 1 watt using POKDYN code because at this level of power no feedback reactivity so important for reactor operation safety. The result of calculation showed that there is no change of significant a period-reactivity relation and transfer function at low power for 2.96 gU/cc, 3.55 gU/cc and 4.8 gU/cc density of silicide fuels. The result of the transfer function at low power showed that the reactor is critical stability with no feedback. The result of calculation also showed that reactivity response no change among three kinds of fuel densities. It can be concluded that from kinetic parameter point of view period-reactivity relation, transfer function at low power, and reactivity response are no change reactor operation from reactivity effect when fuel exchanged. (author)

  11. Solid-Core, Gas-Cooled Reactor for Space and Surface Power

    International Nuclear Information System (INIS)

    King, Jeffrey C.; El-Genk, Mohamed S.

    2006-01-01

    The solid-core, gas-cooled, Submersion-Subcritical Safe Space (S and 4) reactor is developed for future space power applications and avoidance of single point failures. The Mo-14%Re reactor core is loaded with uranium nitride fuel in enclosed cavities, cooled by He-30%Xe, and sized to provide 550 kWth for seven years of equivalent full power operation. The beryllium oxide reflector disassembles upon impact on water or soil. In addition to decreasing the reactor and shadow shield mass, Spectral Shift Absorber (SSA) materials added to the reactor core ensure that it remains subcritical in the worst-case submersion accident. With a 0.1 mm thick boron carbide coating on the outside surface of the core block and 0.25 mm thick iridium sleeves around the fuel stacks, the reflector outer diameter is 43.5 cm and the combined reactor and shadow shield mass is 935.1 kg. With 12.5 atom% gadolinium-155 added to the fuel, 2.0 mm diameter gadolinium-155 sesquioxide intersititial pins, and a 0.1 mm thick gadolinium-155 sesquioxide coating, the S and 4 reactor has a slightly smaller reflector outer diameter of 43.0 cm, and a total reactor and shield mass of 901.7 kg. With 8.0 atom% europium-151 added to the fuel, 2.0 mm diameter europium-151 sesquioxide interstitial pins, and a 0.1 mm thick europium-151 sesquioxide coating, the reflector's outer diameter and the total reactor and shield mass are further reduced to 41.5 cm and 869.2 kg, respectively

  12. Off-gas recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Eppler, M.; Lade, H.J.

    1975-01-01

    According to the invention, it is suggested to provide a buffer vessel in the ring main of the off-gas recirculation system for off-gases of a nuclear reactor to which all chambers or vessels which may contain radioactively contaminated gases are connected, within the connection line to outside air. This is to prevent the immediate release of an appreciable amount of gas to the outside air due to pressure variations conditioned by the sequence of operations - e.g. on the filling of the coolant storage. After the improvement, the released gas may be reduced to the amount of gas corresponding to the leakage gas flow entering the ring mains system. (TK) [de

  13. Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

    Science.gov (United States)

    Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

    2018-02-01

    The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

  14. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  15. Transient thermal-hydraulic simulations of direct cycle gas cooled reactors

    International Nuclear Information System (INIS)

    Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe

    2005-01-01

    This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems

  16. Gas-cooled reactor technology: a bibliography

    International Nuclear Information System (INIS)

    Raleigh, H.D.

    1981-09-01

    Included are 3358 citations on gas-cooled reactor technology contained in the DOE Energy Data Base for the period January 1978 through June 1981. The citations include reports, journal articles, books, conference papers, patents, and monographs. Corporate, Personal Author, Subject, Contract Number, and Report Number Indexes are provided

  17. The world trends of high temperature gas-cooled reactors and the mode of utilization

    International Nuclear Information System (INIS)

    Ishikawa, Hiroshi; Shimokawa, Jun-ichi

    1974-01-01

    After a long period of research and development, high temperature gas-cooled reactors are going to enter the practical stage. The combination of a HTGR with a closed cycle helium gas turbine is advantageous in thermal efficiency, reduction of environmental impact and economy. In recent years, the direct utilization of nuclear heat energy in industries has been attracting interest. The multi-purpose utilization of high temperature gas-cooled reactors is thus now the world trend. Reviewing the world developments in this field, the following matters are described: (1) development of HTGRs in the U.K., West Germany, France and the United States; (2) development of He gas turbine, etc. in West Germany; and (3) multi-purpose utilization of HTGRs in West Germany and Japan. (Mori, K.)

  18. Photosensitized 2-amino-3-hydroxypyridine-induced mitochondrial apoptosis via Smac/DIABLO in human skin cells

    Energy Technology Data Exchange (ETDEWEB)

    Goyal, Shruti; Amar, Saroj Kumar [Photobiology Laboratory, Systems Toxicology and Health Risk Assessment Group, CSIR — Indian Institute of Toxicology Research (CSIR-IITR), M.G, Marg, Lucknow 226001, Uttar Pradesh (India); Academy of Scientific and Innovative Research (AcSIR), CSIR — IITR, Lucknow 226001 (India); Dwivedi, Ashish; Mujtaba, Syed Faiz; Kushwaha, Hari Narayan; Chopra, Deepti; Pal, Manish Kumar; Singh, Dhirendra [Photobiology Laboratory, Systems Toxicology and Health Risk Assessment Group, CSIR — Indian Institute of Toxicology Research (CSIR-IITR), M.G, Marg, Lucknow 226001, Uttar Pradesh (India); Chaturvedi, Rajnish Kumar [Developmental Toxicology Division, CSIR — Indian Institute of Toxicology Research, P. O. Box 80, M.G. Marg, Lucknow 226001 (India); Ray, Ratan Singh, E-mail: ratanray.2011@rediffmail.com [Photobiology Laboratory, Systems Toxicology and Health Risk Assessment Group, CSIR — Indian Institute of Toxicology Research (CSIR-IITR), M.G, Marg, Lucknow 226001, Uttar Pradesh (India); Academy of Scientific and Innovative Research (AcSIR), CSIR — IITR, Lucknow 226001 (India)

    2016-04-15

    The popularity of hair dyes use has been increasing regularly throughout the world as per the demand of hair coloring fashion trends and other cosmetic products. 2-Amino-3-hydroxypyridine (A132) is widely used as a hair dye ingredient around the world. We are reporting first time the phototoxicity mechanism of A132 under ambient environmental UV-B radiation. It showed maximum absorption in UV-B region (317 nm) and forms a photoproduct within an hour exposure of UV-B irradiation. Photocytotoxicity of A132 in human keratinocytes (HaCaT) was measured by mitochondrial (MTT), lysosomal (NRU) and LDH assays which illustrated the significant reduction in cell viability. The role of reactive oxygen species (ROS) generation for A132 phototoxicity was established photo- chemically as well as intracellularly. Noteworthy, formation of tail DNA (comet assay), micronuclei and cyclobutane pyrimidine dimers (CPDs) (immunocytochemistry) formation confirmed the photogenotoxic potential of dye. Cell cycle study (sub-G1peak) and staining with EB/AO revealed the cell cycle arrest and apoptosis. Further, mitochondrial mediated apoptosis was corroborated by reduced MMP, release of cytochrome c and upregulation of caspase-3. Release of mitochondrial Smac/DIABLO in cytoplasm demonstrated the caspase dependent apoptotic cell death by photolabile A132 dye. In-addition increased Bax/Bcl2 ratio again proved the apoptosis. Thus, study suggests that A132 induces photogenotoxicity, phototoxicity and apoptotic cell death through the involvement of Smac/DIABLO in mitochondrial apoptosis via caspase dependent manner. Therefore, the long term use of A132 dye and sunlight exposure jointly increased the oxidative stress in skin which causes premature hair loss, damage to progenitor cells of hair follicles. - Highlights: • Photodegradation of A132 and formation of novel photoproduct • Involvement of ROS in A132 phototoxicity • Role of ROS in DNA damage, CPD and micronuclei formation • Release of

  19. Photosensitized 2-amino-3-hydroxypyridine-induced mitochondrial apoptosis via Smac/DIABLO in human skin cells

    International Nuclear Information System (INIS)

    Goyal, Shruti; Amar, Saroj Kumar; Dwivedi, Ashish; Mujtaba, Syed Faiz; Kushwaha, Hari Narayan; Chopra, Deepti; Pal, Manish Kumar; Singh, Dhirendra; Chaturvedi, Rajnish Kumar; Ray, Ratan Singh

    2016-01-01

    The popularity of hair dyes use has been increasing regularly throughout the world as per the demand of hair coloring fashion trends and other cosmetic products. 2-Amino-3-hydroxypyridine (A132) is widely used as a hair dye ingredient around the world. We are reporting first time the phototoxicity mechanism of A132 under ambient environmental UV-B radiation. It showed maximum absorption in UV-B region (317 nm) and forms a photoproduct within an hour exposure of UV-B irradiation. Photocytotoxicity of A132 in human keratinocytes (HaCaT) was measured by mitochondrial (MTT), lysosomal (NRU) and LDH assays which illustrated the significant reduction in cell viability. The role of reactive oxygen species (ROS) generation for A132 phototoxicity was established photo- chemically as well as intracellularly. Noteworthy, formation of tail DNA (comet assay), micronuclei and cyclobutane pyrimidine dimers (CPDs) (immunocytochemistry) formation confirmed the photogenotoxic potential of dye. Cell cycle study (sub-G1peak) and staining with EB/AO revealed the cell cycle arrest and apoptosis. Further, mitochondrial mediated apoptosis was corroborated by reduced MMP, release of cytochrome c and upregulation of caspase-3. Release of mitochondrial Smac/DIABLO in cytoplasm demonstrated the caspase dependent apoptotic cell death by photolabile A132 dye. In-addition increased Bax/Bcl2 ratio again proved the apoptosis. Thus, study suggests that A132 induces photogenotoxicity, phototoxicity and apoptotic cell death through the involvement of Smac/DIABLO in mitochondrial apoptosis via caspase dependent manner. Therefore, the long term use of A132 dye and sunlight exposure jointly increased the oxidative stress in skin which causes premature hair loss, damage to progenitor cells of hair follicles. - Highlights: • Photodegradation of A132 and formation of novel photoproduct • Involvement of ROS in A132 phototoxicity • Role of ROS in DNA damage, CPD and micronuclei formation • Release of

  20. Steam generator materials constraints in UK design gas-cooled reactors

    International Nuclear Information System (INIS)

    James, D.W.

    1988-01-01

    A widely reported problem with Magnox-type reactors was the oxidation of carbon steel components in gas circuits and steam generators. The effects of temperature, pressure, gas composition and steel composition on oxidation kinetics have been determined, thus allowing the probabilities of failure of critical components to be predicted for a given set of operating conditions. This risk analysis, coupled with regular inspection of reactor and boiler internals, has allowed continued operation of all U.K. Magnox plant. The Advanced Gas Cooled Reactor (AGR) is a direct development of the Magnox design. The first four AGRs commenced operation in 1976, at Hinkley Point 'B' and at Hunterston 'B'. All known materials problems with the steam generators have been diagnosed and solved by the development of appropriate operational strategies, together with minor plant modifications. Materials constraints no longer impose any restrictions to full load performance from the steam generators throughout the predicted life of the plant. Problems discussed in detail are: 1. oxidation of the 9 Cr - 1 Mo superheater. 2. Stress corrosion of the austenitic superheater. 3. Creep of the transition joints between the 9 Cr - 1 Mo and austenitic sections. With the 9 Cr - 1 Mo oxidation maximum temperature restriction virtually removed and creep constraints properly quantified, boiler operation in now favourably placed. Stress corrosion research has allowed the risk of tube failure to be related to time, temperature, stress and chemistry. As a result, the rigorous 'no wetting' policy has been relaxed for the normally high quality AGR feedwater, and the superheat margin has been reduced to 23 deg. C. This has increased the size of the operating window and reduced the number of expensive, and potentially harmful, plant trips. (author)

  1. Russia/Asia-Pacific: rising gas and oil production in the Russian Far East will recast Asia-Pacific energy markets

    International Nuclear Information System (INIS)

    Khartukov, E.

    1994-01-01

    The next few years will see the emergence of the Russian Far East (RFE) as a major new source of hydrocarbons. Expansion of the production of crude oil and refined products will substantially improve the region's degree of energy self-sufficiency. The development of the large gas reserves is likely to have the greatest impact however. Three scenario's for this development are examined. Even under the most pessimistic the region's gas balance will show an exportable surplus by 2000. The energy future of the RFE, especially export plans for gas, depends on foreign investment, though. A radical improvement in the investment climate is needed to promote international co-operation. Political factors and strains within the Russian Federation and dwindling supplies of the crude oil and refined products from Siberia which the RFE still needs, tend to favour the development of economic links between the RFE and its Pacific Rim neighbours. The RFE would then be involved in Pacific energy flows, exporting gas and importing crude oil and refined products to make up its domestic shortfall rather than being dependent on Moscow controlled supplies. Should the RFE take this independent course and open its doors to foreign investors, solutions to the region's energy security problems would be made easier. There would almost certainly, though, be rivalry between the USA, Japan and Korea for influence with the RFE. (3 tables) (UK)

  2. Gas reactor international cooperative program interim report. Pebble bed reactor fuel cycle evaluation

    International Nuclear Information System (INIS)

    1978-09-01

    Nuclear fuel cycles were evaluated for the Pebble Bed Gas Cooled Reactor under development in the Federal Republic of Germany. The basic fuel cycle specified for the HTR-K and PNP is well qualified and will meet the requirements of these reactors. Twenty alternate fuel cycles are described, including high-conversion cycles, net-breeding cycles, and proliferation-resistant cycles. High-conversion cycles, which have a high probability of being successfully developed, promise a significant improvement in resource utilization. Proliferation-resistant cycles, also with a high probability of successful development, compare very favorably with those for other types of reactors. Most of the advanced cycles could be adapted to first-generation pebble bed reactors with no significant modifications

  3. Utilization of multi-purpose high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kawada, Osamu; Onuki, Yoshiaki; Wasaoka, Takeshi.

    1974-01-01

    Concerning the utilization of multi-purpose high temperature gas-cooled reactors, the electric power generation with gas turbines is described: features of HTR-He gas turbine power plants; the state of development of He gas turbines; and combined cycle with gas turbines and steam turbines. The features of gas turbines concern heat dissipation into the environment and the mode of load operation. Outstanding work in the development of He gas turbines is that in Hochtemperatur Helium-Turbine Project in West Germany. The power generation with combined gas turbines and steam turbines appears to be superior to that with gas turbines alone. (Mori, K.)

  4. Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Waloch, M.L.

    1977-01-01

    In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

  5. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  6. Measured gas and particle temperatures in VTT's entrained flow reactor

    DEFF Research Database (Denmark)

    Clausen, Sønnik; Sørensen, L.H.

    2006-01-01

    Particle and gas temperature measurements were carried out in experiments on VTTs entrained flow reactor with 5% and 10% oxygen using Fourier transform infrared emission spectroscopy (FTIR). Particle temperature measurements were performed on polish coal,bark, wood, straw particles, and bark...... and wood particles treated with additive. A two-color technique with subtraction of the background light was used to estimate particle temperatures during experiments. A transmission-emission technique was used tomeasure the gas temperature in the reactor tube. Gas temperature measurements were in good...... agreement with thermocouple readings. Gas lines and bands from CO, CO2 and H2O can be observed in the spectra. CO was only observed at the first measuring port (100ms) with the strongest CO-signal seen during experiments with straw particles. Variations in gas concentration (CO2 and H2O) and the signal from...

  7. The modular high-temperature gas-cooled reactor - a new production reactor

    International Nuclear Information System (INIS)

    Nulton, J.D.

    1990-01-01

    One of the reactor concepts being considered for application as a new production reactor (NPR) is a 350-MW(thermal) modular high-temperature gas-cooled reactor (MHTGR). The proposed MHTGR-NPR is based on the design of the commercial MHTGR and is being developed by a team that includes General Atomics and Combustion Engineering. The proposed design includes four modules combined into a production block that includes a shared containment, a spent-fuel storage facility, and other support facilities. The MHTGR has a helium-cooled, graphite-moderated, graphite-reflected annular core formed from prismatic graphite fuel blocks. The MHTGR fuel consists of highly enriched uranium oxycarbide (UCO) microsphere fuel particles that are coated with successive layers of pyrolytic carbon (PyC) and silicon carbide (SiC). Tritium-producing targets consist of enriched 6 Li aluminate microsphere target particles that are coated with successive layers of PyC and SiC similar to the fuel microspheres. Normal reactivity control is implemented by articulated control rods that can be inserted into channels in the inner and outer reflector blocks. Shutdown heat removal is accomplished by a single shutdown heat exchanger and electric motor-driven circulator located in the bottom of the reactor vessel. Current plans are to stack spent fuel elements in dry, helium-filled, water-cooled wells and store them for ∼1 yr before reprocessing. All phases of MHTGR fuel reprocessing have been demonstrated

  8. Late Holocene sea ice conditions in Herald Canyon, Chukchi Sea

    Science.gov (United States)

    Pearce, C.; O'Regan, M.; Rattray, J. E.; Hutchinson, D. K.; Cronin, T. M.; Gemery, L.; Barrientos, N.; Coxall, H.; Smittenberg, R.; Semiletov, I. P.; Jakobsson, M.

    2017-12-01

    Sea ice in the Arctic Ocean has been in steady decline in recent decades and, based on satellite data, the retreat is most pronounced in the Chukchi and Beaufort seas. Historical observations suggest that the recent changes were unprecedented during the last 150 years, but for a longer time perspective, we rely on the geological record. For this study, we analyzed sediment samples from two piston cores from Herald Canyon in the Chukchi Sea, collected during the 2014 SWERUS-C3 Arctic Ocean Expedition. The Herald Canyon is a local depression across the Chukchi Shelf, and acts as one of the main pathways for Pacific Water to the Arctic Ocean after entering through the narrow and shallow Bering Strait. The study site lies at the modern-day seasonal sea ice minimum edge, and is thus an ideal location for the reconstruction of past sea ice variability. Both sediment cores contain late Holocene deposits characterized by high sediment accumulation rates (100-300 cm/kyr). Core 2-PC1 from the shallow canyon flank (57 m water depth) is 8 meter long and extends back to 4200 cal yrs BP, while the upper 3 meters of Core 4-PC1 from the central canyon (120 mwd) cover the last 3000 years. The chronologies of the cores are based on radiocarbon dates and the 3.6 ka Aniakchak CFE II tephra, which is used as an absolute age marker to calculate the marine radiocarbon reservoir age. Analysis of biomarkers for sea ice and surface water productivity indicate stable sea ice conditions throughout the entire late Holocene, ending with an abrupt increase of phytoplankton sterols in the very top of both sediment sequences. The shift is accompanied by a sudden increase in coarse sediments (> 125 µm) and a minor change in δ13Corg. We interpret this transition in the top sediments as a community turnover in primary producers from sea ice to open water biota. Most importantly, our results indicate that the ongoing rapid ice retreat in the Chukchi Sea of recent decades was unprecedented during the

  9. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

  10. Improvements in gas supply systems for heavy-water moderated reactors

    International Nuclear Information System (INIS)

    Aubert, G.; Hassig, J.M.; Laurent, N.; Thomas, B.

    1964-01-01

    In a heavy-water moderated reactor cooled by pressurized gas, an important problem from the point of view, of the reactor block and its economics is the choice of the gas supply system. In the pressure tube solution, the whole of the reactor block structure is at a relatively low temperature, whereas the gas supply equipment is at that of the gas, which is much higher. These parts, through which passes the heat carrying fluid have to present as low a resistance as possible to it so as to avoid costly extra blowing power. Finally, they may only be placed in the reactor block after it has been built; the time required for putting them in position should therefore not be too long. The work reported here concerns the various problems arising in the case of each channel being supplied individually by a tube at the entry and the exit which is connected to a main circuit made up of large size collectors. This individual tubing is sufficiently flexible to absorb the differential expansion and the movement of its ends without stresses or prohibitive reactions being produced; the tubing is also of relatively short length so as to reduce the pressure head of the pressurized gas outside the channels; the small amount of space taken up by the tubing makes it possible to assemble it in a manner which is satisfactory from the point of view both of the time required and of the technical quality. (authors) [fr

  11. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  12. Tectonic activity and the evolution of submarine canyons: The Cook Strait Canyon system, New Zealand

    Science.gov (United States)

    Micallef, Aaron; Mountjoy, Joshu; Barnes, Philip; Canals, Miquel; Lastras, Galderic

    2016-04-01

    Submarine canyons are Earth's most dramatic erosional features, comprising steep-walled valleys that originate in the continental shelf and slope. They play a key role in the evolution of continental margins by transferring sediments into deep water settings and are considered important biodiversity hotspots, pathways for nutrients and pollutants, and analogues of hydrocarbon reservoirs. Although comprising only one third of continental margins worldwide, active margins host more than half of global submarine canyons. We still lack of thorough understanding of the coupling between active tectonics and submarine canyon processes, which is necessary to improve the modelling of canyon evolution in active margins and derive tectonic information from canyon morphology. The objectives of this study are to: (i) understand how tectonic activity influences submarine canyon morphology, processes, and evolution in an active margin, and (2) formulate a generalised model of canyon development in response to tectonic forcing based on morphometric parameters. We fulfil these objectives by analysing high resolution geophysical data and imagery from Cook Strait Canyon system, offshore New Zealand. Using these data, we demonstrate that tectonic activity, in the form of major faults and structurally-generated tectonic ridges, leaves a clear topographic signature on submarine canyon location and morphology, in particular their dendritic and sinuous planform shapes, steep and linear longitudinal profiles, and cross-sectional asymmetry and width. We also report breaks/changes in canyon longitudinal slope gradient, relief and slope-area regression models at the intersection with faults. Tectonic activity gives rise to two types of knickpoints in the Cook Strait Canyon. The first type consists of low slope gradient, rounded and diffusive knickpoints forming as a result of short wavelength folds or fault break outs and being restored to an equilibrium profile by upstream erosion and

  13. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  14. The early history of high-temperature helium gas-cooled nuclear power reactors

    International Nuclear Information System (INIS)

    Simnad, M.T.; California Univ., San Diego, La Jolla, CA

    1991-01-01

    The original concepts in the proposals for high-temperature helium gas-cooled power reactors by Farrington Daniels, during the decade 1944-1955, are summarized. The early research on the development of the helium gas-cooled power reactors is reviewed, and the operational experiences with the first generation of HTGRs are discussed. (author)

  15. First Study of Helium Gas Purification System as Primary Coolant of Co-Generation Reactor

    International Nuclear Information System (INIS)

    Piping Supriatna

    2009-01-01

    The technological progress of NPP Generation-I on 1950’s, Generation-II, Generation-III recently on going, and Generation-IV which will be implemented on next year 2025, concept of nuclear power technology implementation not only for generate electrical energy, but also for other application which called cogeneration reactor. Commonly the type of this reactor is High Temperature Reactor (HTR), which have other capabilities like Hydrogen production, desalination, Enhanced Oil Recovery (EOR), etc. The cogeneration reactor (HTR) produce thermal output higher than commonly Nuclear Power Plant, and need special Heat Exchanger with helium gas as coolant. In order to preserve heat transfer with high efficiency, constant purity of the gas must be maintained as well as possible, especially contamination from its impurities. In this report has been done study for design concept of HTR primary coolant gas purification system, including methodology by sampling He gas from Primary Coolant and purification by using Physical Helium Splitting Membrane. The examination has been designed in physical simulator by using heater as reactor core. The result of study show that the of Primary Coolant Gas Purification System is enable to be implemented on cogeneration reactor. (author)

  16. 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)

  17. Experimental investigation of a pilot-scale jet bubbling reactor for wet flue gas desulphurisation

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Kiil, Søren; Johnsson, Jan Erik

    2003-01-01

    In the present work, an experimental parameter study was conducted in a pilot-scale jet bubbling reactor for wet flue gas desulphurisation (FGD). The pilot plant is downscaled from a limestone-based, gypsum producing full-scale wet FGD plant. Important process parameters, such as slurry pH, inlet...... flue gas concentration of SO2, reactor temperature, and slurry concentration of Cl- have been varied. The degree of desulphurisation, residual limestone content of the gypsum, liquid phase concentrations, and solids content of the slurry were measured during the experimental series. The SO2 removal...... efficiency increased from 66.1% to 71.5% when the reactor slurry pH was changed from 3.5 to 5.5. Addition of Cl(in the form of CaCl2 . 2H(2)O) to the slurry (25 g Cl-/l) increased the degree of desulphurisation to above 99%, due to the onset of extensive foaming, which substantially increased the gas...

  18. 75 FR 54618 - CAlifornians for Renewable Energy, Inc. (CARE) v. Pacific Gas and Electric Company, Southern...

    Science.gov (United States)

    2010-09-08

    ... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. EL10-84-000] CAlifornians for Renewable Energy, Inc. (CARE) v. Pacific Gas and Electric Company, Southern California Edison Company, San Diego Gas & Electric Company, California Public Utilities Commission; Notice of Complaint...

  19. 75 FR 66744 - Californians for Renewable Energy, Inc. (CARE) v. Pacific Gas and Electric Company, Southern...

    Science.gov (United States)

    2010-10-29

    ... DEPARTMENT OF ENERGY Federal Energy Regulatory Commission [Docket No. EL10-84-001] Californians for Renewable Energy, Inc. (CARE) v. Pacific Gas and Electric Company, Southern California Edison Company, San Diego Gas & Electric Company, California Public Utilities Commission; Notice of Amended...

  20. Fueling moving ring field-reversed mirror reactor plasmas

    International Nuclear Information System (INIS)

    Felber, F.S.

    1980-01-01

    The concept of small fusion reactors is being studied jointly by Lawrence Livermore Laboratory General Atomic Company, and Pacific Gas and Electric Company. The objective is to investigate alternatives and then to develop a conceptual design for a small reactor that could produce useful, though not necessarily economical, energy by the late 1980s. Three methods of fueling a small moving ring field-reversed mirror are considered: injection of fuel pellets accelerated by laser ablation, injection of fuel pellets accelerated by deflagration-gun ablation, and direct injection of plasma by a deflagration gun. 13 refs

  1. Radiological considerations of the reactor cover gas processing system at the FFTF

    International Nuclear Information System (INIS)

    Prevo, P.R.

    1987-01-01

    Radiological and environmental protection experience associated with the reactor cover gas processing system at the Fast Flux Test Facility (FFTF) has been excellent. Personnel radiation exposures received from operating and maintaining the reactor cover gas processing system have been very low, the system has remained free of radioactive particulate contamination through the first seven operating cycles (cesium contamination was detected at the end of Cycle 8A), and releases of radioactivity to the environment have been very low, well below environmental standards. This report discusses these three aspects of fast reactor cover gas purification over the first eight operating cycles of the FFTF (a duration of a little more than four years, from April 1982 through July 1986). (author)

  2. Gas pollutant cleaning by a membrane reactor

    Directory of Open Access Journals (Sweden)

    Kaldis Sotiris

    2006-01-01

    Full Text Available An alternative technology for the removal of gas pollutants at the integrated gasification combined cycle process for power generation is the use of a catalytic membrane reactor. In the present study, ammonia decomposition in a catalytic reactor, with a simultaneous removal of hydrogen through a ceramic membrane, was investigated. A Ni/Al2O3 catalyst was prepared by the dry and wet impregnation method and characterized by the inductively coupled plasma method, scanning electron microscopy, X-ray diffraction, and N2 adsorption before and after activation. Commercially available a-Al2O3 membranes were also characterized and the permeabilities and permselectivities of H2, N2, and CO2 were measured by the variable volume method. In parallel with the experimental analysis, the necessary mathematical models were developed to describe the operation of the catalytic membrane reactor and to compare its performance with the conventional reactor. .

  3. Calcium oxide/carbon dioxide reactivity in a packed bed reactor of a chemical heat pump for high-temperature gas reactors

    International Nuclear Information System (INIS)

    Kato, Yukitaka; Yamada, Mitsuteru; Kanie, Toshihiro; Yoshizawa, Yoshio

    2001-01-01

    The thermal performance of a chemical heat pump that uses a calcium oxide/carbon dioxide reaction system was discussed as a heat storage system for utilizing heat output from high temperature gas reactors (HTGR). Calcium oxide/carbon dioxide reactivity for the heat pump was measured using a packed bed reactor containing 1.0 kg of reactant. The reactor was capable of storing heat at 900 deg. C by decarbonation of calcium carbonate and generating up to 997 deg. C by carbonation of calcium oxide. The amount of stored heat in the reactor was 800-900 kJ kg -1 . The output temperature of the reactor could be controlled by regulating the carbonation pressure. The thermal storage performance of the reactor was superior to that of conventional sensible heat storage systems. A heat pump using this CaO/CO 2 reactor is expected to contribute to thermal load leveling and to realize highly efficient utilization of HTGR output due to the high heat storage density and high-quality temperature output of the heat pump

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

    Directory of Open Access Journals (Sweden)

    Lei Zhang

    2018-05-01

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

  5. Evaluation of the Gas Turbine Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs.

  6. Evaluation of the Gas Turbine Modular Helium Reactor

    International Nuclear Information System (INIS)

    1994-02-01

    Recent advances in gas-turbine and heat exchanger technology have enhanced the potential for a Modular Helium Reactor (MHR) incorporating a direct gas turbine (Brayton) cycle for power conversion. The resulting Gas Turbine Modular Helium Reactor (GT-MHR) power plant combines the high temperature capabilities of the MHR with the efficiency and reliability of modern gas turbines. While the passive safety features of the steam cycle MHR (SC-MHR) are retained, generation efficiencies are projected to be in the range of 48% and steam power conversion systems, with their attendant complexities, are eliminated. Power costs are projected to be reduced by about 20%, relative to the SC-MHR or coal. This report documents the second, and final, phase of a two-part evaluation that concluded with a unanimous recommendation that the direct cycle (DC) variant of the GT-MHR be established as the commercial objective of the US Gas-Cooled Reactor Program. This recommendation has been endorsed by industrial and utility participants and accepted by the US Department of Energy (DOE). The Phase II effort, documented herein, concluded that the DC GT-MHR offers substantial technical and economic advantages over both the IDC and SC systems. Both the DC and IDC were found to offer safety advantages, relative to the SC, due to elimination of the potential for water ingress during power operations. This is the dominant consequence event for the SC. The IDC was judged to require somewhat less development than the direct cycle, while the SC, which has the greatest technology base, incurs the least development cost and risk. While the technical and licensing requirements for the DC were more demanding, they were judged to be incremental and feasible. Moreover, the DC offers significant performance and cost improvements over the other two concepts. Overall, the latter were found to justify the additional development needs

  7. Geomorphic process fingerprints in submarine canyons

    Science.gov (United States)

    Brothers, Daniel S.; ten Brink, Uri S.; Andrews, Brian D.; Chaytor, Jason D.; Twichell, David C.

    2013-01-01

    Submarine canyons are common features of continental margins worldwide. They are conduits that funnel vast quantities of sediment from the continents to the deep sea. Though it is known that submarine canyons form primarily from erosion induced by submarine sediment flows, we currently lack quantitative, empirically based expressions that describe the morphology of submarine canyon networks. Multibeam bathymetry data along the entire passive US Atlantic margin (USAM) and along the active central California margin near Monterey Bay provide an opportunity to examine the fine-scale morphology of 171 slope-sourced canyons. Log–log regression analyses of canyon thalweg gradient (S) versus up-canyon catchment area (A) are used to examine linkages between morphological domains and the generation and evolution of submarine sediment flows. For example, canyon reaches of the upper continental slope are characterized by steep, linear and/or convex longitudinal profiles, whereas reaches farther down canyon have distinctly concave longitudinal profiles. The transition between these geomorphic domains is inferred to represent the downslope transformation of debris flows into erosive, canyon-flushing turbidity flows. Over geologic timescales this process appears to leave behind a predictable geomorphic fingerprint that is dependent on the catchment area of the canyon head. Catchment area, in turn, may be a proxy for the volume of sediment released during geomorphically significant failures along the upper continental slope. Focused studies of slope-sourced submarine canyons may provide new insights into the relationships between fine-scale canyon morphology and down-canyon changes in sediment flow dynamics.

  8. Natural gas network resiliency to a "shakeout scenario" earthquake.

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, James F.; Corbet, Thomas Frank,; Brooks, Robert E.

    2013-06-01

    A natural gas network model was used to assess the likely impact of a scenario San Andreas Fault earthquake on the natural gas network. Two disruption scenarios were examined. The more extensive damage scenario assumes the disruption of all three major corridors bringing gas into southern California. If withdrawals from the Aliso Canyon storage facility are limited to keep the amount of stored gas within historical levels, the disruption reduces Los Angeles Basin gas supplies by 50%. If Aliso Canyon withdrawals are only constrained by the physical capacity of the storage system to withdraw gas, the shortfall is reduced to 25%. This result suggests that it is important for stakeholders to put agreements in place facilitating the withdrawal of Aliso Canyon gas in the event of an emergency.

  9. Reducing errors in aircraft atmospheric inversion estimates of point-source emissions: the Aliso Canyon natural gas leak as a natural tracer experiment

    Science.gov (United States)

    Gourdji, S. M.; Yadav, V.; Karion, A.; Mueller, K. L.; Conley, S.; Ryerson, T.; Nehrkorn, T.; Kort, E. A.

    2018-04-01

    Urban greenhouse gas (GHG) flux estimation with atmospheric measurements and modeling, i.e. the ‘top-down’ approach, can potentially support GHG emission reduction policies by assessing trends in surface fluxes and detecting anomalies from bottom-up inventories. Aircraft-collected GHG observations also have the potential to help quantify point-source emissions that may not be adequately sampled by fixed surface tower-based atmospheric observing systems. Here, we estimate CH4 emissions from a known point source, the Aliso Canyon natural gas leak in Los Angeles, CA from October 2015–February 2016, using atmospheric inverse models with airborne CH4 observations from twelve flights ≈4 km downwind of the leak and surface sensitivities from a mesoscale atmospheric transport model. This leak event has been well-quantified previously using various methods by the California Air Resources Board, thereby providing high confidence in the mass-balance leak rate estimates of (Conley et al 2016), used here for comparison to inversion results. Inversions with an optimal setup are shown to provide estimates of the leak magnitude, on average, within a third of the mass balance values, with remaining errors in estimated leak rates predominantly explained by modeled wind speed errors of up to 10 m s‑1, quantified by comparing airborne meteorological observations with modeled values along the flight track. An inversion setup using scaled observational wind speed errors in the model-data mismatch covariance matrix is shown to significantly reduce the influence of transport model errors on spatial patterns and estimated leak rates from the inversions. In sum, this study takes advantage of a natural tracer release experiment (i.e. the Aliso Canyon natural gas leak) to identify effective approaches for reducing the influence of transport model error on atmospheric inversions of point-source emissions, while suggesting future potential for integrating surface tower and

  10. Nuclear power for coexistence with nature, high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kaneko, Yoshihiko

    1996-01-01

    Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

  11. Anatomy of La Jolla submarine canyon system; offshore southern California

    Science.gov (United States)

    Paull, C.K.; Caress, D.W.; Lundsten, E.; Gwiazda, R.; Anderson, K.; McGann, M.; Conrad, J.; Edwards, B.; Sumner, E.J.

    2013-01-01

    An autonomous underwater vehicle (AUV) carrying a multibeam sonar and a chirp profiler was used to map sections of the seafloor within the La Jolla Canyon, offshore southern California, at sub-meter scales. Close-up observations and sampling were conducted during remotely operated vehicle (ROV) dives. Minisparker seismic-reflection profiles from a surface ship help to define the overall geometry of the La Jolla Canyon especially with respect to the pre-canyon host sediments. The floor of the axial channel is covered with unconsolidated sand similar to the sand on the shelf near the canyon head, lacks outcrops of the pre-canyon host strata, has an almost constant slope of 1.0° and is covered with trains of crescent shaped bedforms. The presence of modern plant material entombed within these sands confirms that the axial channel is presently active. The sand on the canyon floor liquefied during vibracore collection and flowed downslope, illustrating that the sediment filling the channel can easily fail even on this gentle slope. Data from the canyon walls help constrain the age of the canyon and extent of incision. Horizontal beds of moderately cohesive fine-grained sediments exposed on the steep canyon walls are consistently less than 1.232 million years old. The lateral continuity of seismic reflectors in minisparker profiles indicate that pre-canyon host strata extend uninterrupted from outside the canyon underneath some terraces within the canyon. Evidence of abandoned channels and point bar-like deposits are noticeably absent on the inside bend of channel meanders and in the subsurface of the terraces. While vibracores from the surface of terraces contain thin (art seafloor mapping and exploration tools provides a uniquely detailed view of the morphology within an active submarine canyon.

  12. Emergency cooling system for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Cook, R.K.; Burylo, P.S.

    1975-01-01

    The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

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

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Tarigan, Alim

    1999-01-01

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

  15. Environmental Assessment for East Housing Area Solar Energy Project, Vandenberg Air Force Base, California

    Science.gov (United States)

    2014-07-01

    May. Gevirtz, Elihu, 2014. Email communication with Craig Woodman dated 20 May, 2014. Glassow, Michael A. 1996. Purisimeño Chumash Prehistory ...William C. Sturtevant, general editor. Smithsonian Institution, Washington, D.C. 1972. 9000 Years of Prehistory at Diablo Canyon, San Luis Obispo

  16. Thermodynamic performance of a gas-core fission reactor

    International Nuclear Information System (INIS)

    Klein, W.

    1987-01-01

    The purpose of this thesis was to investigate the thermodynamic behaviour of a critical quantity of gaseous uranium-fluorides in chemical equilibrium with a graphite wall. From the very beginning a container was considered with cooled walls. As it was evident that a nuclear reactor working with gaseous fuel should run at much higher temperatures than classical LWR or HTGR reactors, most of the investigations were performed for walls with a surface temperature of 1800 to 2000 K. It was supposed that such a surface temperature would be technologically possible for a heat load between 1 and 5 MWatt m -2 . Cooling with high pressure helium-gas has to keep balance with this heat flux. The technical construction of such a wall will be a problem in itself. It is thought that the experiences with re-entry-vessels in space-technology can be used. A basic assumption in all the calculations is that the U-C-F reactor gas 'sees' a graphite wall, possibly graphite tiles supported by heat resistant materials like SiN 2 , SiC 2 and at a lower temperature level by niobium-steel. Such a gastight compound-system is not necessarily of high-tensile strength materials. It has to be surrounded by a cooled neutron moderator-reflector which in its turn must be supported by a steel-wall at room temperature holding pressure of the order of 100 bar (10 MPa). The design of such a compound-wall is a task for the future. 116 refs.; 28 figs.; 29 tabs

  17. The real gas behaviour of helium as a cooling medium for high-temperature reactors

    International Nuclear Information System (INIS)

    Hewing, G.

    1977-01-01

    The article describes the influence of the real gas behaviour on the variables of state for the helium gas and the effects on the design of high-temperature reactor plants. After explaining the basic equations for describing variables and changes of state of the real gas, the real and ideal gas behaviour is analysed. Finally, the influence of the real gas behaviour on the design of high-temperature reactors in one- and two-cycle plants is investigated. (orig.) [de

  18. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  19. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  20. Flow in bedrock canyons.

    Science.gov (United States)

    Venditti, Jeremy G; Rennie, Colin D; Bomhof, James; Bradley, Ryan W; Little, Malcolm; Church, Michael

    2014-09-25

    Bedrock erosion in rivers sets the pace of landscape evolution, influences the evolution of orogens and determines the size, shape and relief of mountains. A variety of models link fluid flow and sediment transport processes to bedrock incision in canyons. The model components that represent sediment transport processes are increasingly well developed. In contrast, the model components being used to represent fluid flow are largely untested because there are no observations of the flow structure in bedrock canyons. Here we present a 524-kilometre, continuous centreline, acoustic Doppler current profiler survey of the Fraser Canyon in western Canada, which includes 42 individual bedrock canyons. Our observations of three-dimensional flow structure reveal that, as water enters the canyons, a high-velocity core follows the bed surface, causing a velocity inversion (high velocities near the bed and low velocities at the surface). The plunging water then upwells along the canyon walls, resulting in counter-rotating, along-stream coherent flow structures that diverge near the bed. The resulting flow structure promotes deep scour in the bedrock channel floor and undercutting of the canyon walls. This provides a mechanism for channel widening and ensures that the base of the walls is swept clear of the debris that is often deposited there, keeping the walls nearly vertical. These observations reveal that the flow structure in bedrock canyons is more complex than assumed in the models presently used. Fluid flow models that capture the essence of the three-dimensional flow field, using simple phenomenological rules that are computationally tractable, are required to capture the dynamic coupling between flow, bedrock erosion and solid-Earth dynamics.

  1. State of development of high temperature gas-cooled reactors in foreign countries

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

    Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

  2. Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

    International Nuclear Information System (INIS)

    Bratton, R.L.

    1997-05-01

    As a part of the Pacific Northwest National Laboratory's Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations

  3. Development of the IAEA’s Knowledge Preservation Portals for Fast Reactors and Gas-Cooled Reactors Knowledge Preservation

    International Nuclear Information System (INIS)

    Batra, C.; Menahem, D. Beraha; Kriventsev, V.; Monti, S.; Reitsma, F.; Grosbois, J. de; Khoroshev, M.; Gladyshev, M.

    2016-01-01

    Full text: The IAEA has been carrying out a dedicated initiative on fast reactor knowledge preservation since 2003. The main objectives of the Fast Reactor Knowledge Portal (FRKP) initiative are to, a) halt the on-going loss of information related to fast reactors (FR), and b) collect, retrieve, preserve and make accessible existing data and information on FR. This portal will help in knowledge sharing, development, search and discovery, collaboration and communication of fast reactor related information. On similar lines a Gas Cooled Fast Reactor Knowledge Preservation portal project also started in 2013. Knowledge portals are capable to control and manage both publicly available as well as controlled information. The portals will not only incorporate existing set of knowledge and information, but will also provide a systemic platform for further preservation of new developments. It will include fast reactor and gas cooled reactor document repositories, project workspaces for the IAEA’s Coordinated Research Projects (CRPs), Technical Meetings (TMs), forums for discussion, etc. The portal will also integrate a taxonomy based search tool, which will help using new semantic search capabilities for improved conceptual retrieve of documents. The taxonomy complies with international web standards as defined by the W3C (World Wide Web Consortium). (author

  4. Anatomy of La Jolla submarine canyon system; offshore southern California

    Science.gov (United States)

    Paull, C.K.; Caress, D.W.; Lundsten, E.; Gwiazda, R.; Anderson, K.; McGann, M.; Conrad, J.; Edwards, B.; Sumner, E.J.

    2013-01-01

    An autonomous underwater vehicle (AUV) carrying a multibeam sonar and a chirp profiler was used to map sections of the seafloor within the La Jolla Canyon, offshore southern California, at sub-meter scales. Close-up observations and sampling were conducted during remotely operated vehicle (ROV) dives. Minisparker seismic-reflection profiles from a surface ship help to define the overall geometry of the La Jolla Canyon especially with respect to the pre-canyon host sediments. The floor of the axial channel is covered with unconsolidated sand similar to the sand on the shelf near the canyon head, lacks outcrops of the pre-canyon host strata, has an almost constant slope of 1.0° and is covered with trains of crescent shaped bedforms. The presence of modern plant material entombed within these sands confirms that the axial channel is presently active. The sand on the canyon floor liquefied during vibracore collection and flowed downslope, illustrating that the sediment filling the channel can easily fail even on this gentle slope. Data from the canyon walls help constrain the age of the canyon and extent of incision. Horizontal beds of moderately cohesive fine-grained sediments exposed on the steep canyon walls are consistently less than 1.232 million years old. The lateral continuity of seismic reflectors in minisparker profiles indicate that pre-canyon host strata extend uninterrupted from outside the canyon underneath some terraces within the canyon. Evidence of abandoned channels and point bar-like deposits are noticeably absent on the inside bend of channel meanders and in the subsurface of the terraces. While vibracores from the surface of terraces contain thin (< 10 cm) turbidites, they are inferred to be part of a veneer of recent sediment covering pre-canyon host sediments that underpin the terraces. The combined use of state of the art seafloor mapping and exploration tools provides a uniquely detailed view of the morphology within an active submarine canyon.

  5. Utilization of research reactors - A global perspective

    International Nuclear Information System (INIS)

    Muranaka, R.G.

    1988-01-01

    This paper presents 1) a worldwide picture of research reactors, operable, shutdown, under construction and planned, 2) statistics on utilization of research reactors including TRIGA reactors, and 3) some results of a survey conducted during 1988 on the utilization of research reactors in developing Member States in the Asia-Pacific Region

  6. Acoustical environment of gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Blevins, R.D.

    1986-01-01

    Methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound, are described. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed. (author)

  7. Specialists' meeting on fast reactor cover gas purification

    International Nuclear Information System (INIS)

    1987-01-01

    The tentative agenda was adopted by the participants without comment and was followed throughout the meeting. The following topics were discussed at the subsequent sessions of the meeting on 'Fast Reactor Cover Gas Purification': National Position Papers; Impurities: Sources and Measurement; Cover Gas Purification Techniques; Sodium Aerosol Trapping; Radiological Considerations. Based on the papers presented and the discussions following, session summaries and conclusions were prepared and are included in this report

  8. Specialists' meeting on fast reactor cover gas purification

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1987-07-01

    The tentative agenda was adopted by the participants without comment and was followed throughout the meeting. The following topics were discussed at the subsequent sessions of the meeting on 'Fast Reactor Cover Gas Purification': National Position Papers; Impurities: Sources and Measurement; Cover Gas Purification Techniques; Sodium Aerosol Trapping; Radiological Considerations. Based on the papers presented and the discussions following, session summaries and conclusions were prepared and are included in this report.

  9. Transience and persistence of natural hydrocarbon seepage in Mississippi Canyon, Gulf of Mexico

    Science.gov (United States)

    Garcia-Pineda, Oscar; MacDonald, Ian; Silva, Mauricio; Shedd, William; Daneshgar Asl, Samira; Schumaker, Bonny

    2016-07-01

    Analysis of the magnitude of oil discharged from natural hydrocarbon seeps can improve understanding of the carbon cycle and the Gulf of Mexico (GOM) ecosystem. With use of a large archive of remote sensing data, in combination with geophysical and multibeam data, we identified, mapped, and characterized natural hydrocarbon seeps in the Macondo prospect region near the wreck site of the drill-rig Deepwater Horizon (DWH). Satellite image processing and the cluster analysis revealed locations of previously undetected seep zones. Including duplicate detections, a total of 562 individual gas plumes were also observed in multibeam surveys. In total, SAR imagery confirmed 52 oil-producing seep zones in the study area. In almost all cases gas plumes were associated with oil-producing seep zones. The cluster of seeps in the vicinity of lease block MC302 appeared to host the most persistent and prolific oil vents. Oil slicks and gas plumes observed over the DWH site were consistent with discharges of residual oil from the wreckage. In contrast with highly persistent oil seeps observed in the Green Canyon and Garden Banks lease areas, the seeps in the vicinity of Macondo Prospect were intermittent. The difference in the number of seeps and the quantity of surface oil detected in Green Canyon was almost two orders of magnitude greater than in Mississippi Canyon.

  10. Study of new structures adapted to gas-graphite and gas-heavy water reactors

    International Nuclear Information System (INIS)

    Martin, R.; Roche, R.

    1964-01-01

    The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [fr

  11. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  12. Gas-cooled reactor power systems for space

    International Nuclear Information System (INIS)

    Walter, C.E.

    1987-01-01

    Efficiency and mass characteristics for four gas-cooled reactor power system configurations in the 2- to 20-MWe power range are modeled. The configurations use direct and indirect Brayton cycles with and without regeneration in the power conversion loop. The prismatic ceramic core of the reactor consists of several thousand pencil-shaped tubes made from a homogeneous mixture of moderator and fuel. The heat rejection system is found to be the major contributor to system mass, particularly at high power levels. A direct, regenerated Brayton cycle with helium working fluid permits high efficiency and low specific mass for a 10-MWe system

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

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1980-05-01

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

  14. The evolution of gasification processes and reactors and the utilization of the coal gas. A proposition for the implementation of the gasification technology

    International Nuclear Information System (INIS)

    Pasculete, E.; Iorgulescu, S.

    1996-01-01

    Thermochemical treatment of coal by gasification, considered as a non-polluting technology to turn the coal highly-profitably is one of the alternative ways to produce gas with a high effective caloric capacity. Due to its advantages, the gasification has made through the last few decades significant advances from the point of view of the process efficiency (chemical, thermal), of motor outputs (in m 3 producer gas / m 2 reactor cross section x hour), of the solutions of supplying energy to support the endothermic reactions implied by the process, and especially of the reactors. Reactors have been developed from gas generators. Starting from gas generators various advanced reactors (of 1 st to 3 rd generation) have been developed to produce air gas, water gas or mixed gas. Applications of the producer gas were developed using it either as fuel or as synthesis gas in chemical industry or else as a substitute to the natural gas in combined cycle gas turbines where the gasification plant was integrated. In Romania there are projects in the field of coal gasification, namely at ICPET-RESEARCH, that can offer advanced technologies. One of these projects deals with the construction of the first demonstrative gasification plant based on a highly efficient process and equipped with a 10 G cal/h reactor. (author). 1 tab., 12 refs

  15. Carbonaceous aerosol particles from common vegetation in the Grand Canyon

    International Nuclear Information System (INIS)

    Hallock, K.A.; Mazurek, M.A.; Cass, G.R.

    1992-05-01

    The problem of visibility reduction in the Grand Canyon due to fine organic aerosol particles in the atmosphere has become an area of increased environmental concern. Aerosol particles can be derived from many emission sources. In this report, we focus on identifying organic aerosols derived from common vegetation in the Grand Canyon. These aerosols are expected to be significant contributors to the total atmospheric organic aerosol content. Aerosol samples from living vegetation were collected by resuspension of surface wax and resin components liberated from the leaves of vegetation common to areas of the Grand Canyon. The samples were analyzed using high-resolution gas chromatography/mass spectrometry (GC/MS). Probable identification of compounds was made by comparison of sample spectra with National Institute of Standards and Technology (NIST) mass spectral references and positive identification of compounds was made when possible by comparison with authentic standards as well as NIST references. Using these references, we have been able to positively identify the presence of n-alkane and n-alkanoic acid homolog series in the surface waxes of the vegetation sampled. Several monoterpenes, sesquiterpenes, and diterpenes were identified also as possible biogenic aerosols which may contribute to the total organic aerosol abundance leading to visibility reduction in the Grand Canyon

  16. Steam conversion of liquefied petroleum gas and methane in microchannel reactor

    Science.gov (United States)

    Dimov, S. V.; Gasenko, O. A.; Fokin, M. I.; Kuznetsov, V. V.

    2018-03-01

    This study presents experimental results of steam conversion of liquefied petroleum gas and methane in annular catalytic reactor - heat exchanger. The steam reforming was done on the Rh/Al2O3 nanocatalyst with the heat applied through the microchannel gap from the outer wall. Concentrations of the products of chemical reactions in the outlet gas mixture are measured at different temperatures of reactor. The range of channel wall temperatures at which the ratio of hydrogen and carbon oxide in the outlet mixture grows substantially is determined. Data on the composition of liquefied petroleum gas conversion products for the ratio S/C = 5 was received for different GHVS.

  17. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  18. Fuel assembly for gas-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.

    1976-01-01

    A fuel assembly is described for gas-cooled nuclear reactor which consists of a wrapper tube within which are positioned a number of spaced apart beds in a stack, with each bed containing spherical coated particles of fuel; each of the beds has a perforated top and bottom plate; gaseous coolant passes successively through each of the beds; through each of the beds also passes a bypass tube; part of the gas travels through the bed and part passes through the bypass tube; the gas coolant which passes through both the bed and the bypass tube mixes in the space on the outlet side of the bed before entering the next bed

  19. Fuel arrangement for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Tobin, J.M.

    1978-01-01

    Disclosed is a fuel arrangement for a high temperature gas cooled reactor including fuel assemblies with separate directly cooled fissile and fertile fuel elements removably inserted in an elongated moderator block also having a passageway for control elements

  20. Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

    International Nuclear Information System (INIS)

    Halvers, L.J.

    1977-03-01

    An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

  1. Impact of roof height non-uniformity on pollutant transport between a street canyon and intersections.

    Science.gov (United States)

    Nosek, Štěpán; Kukačka, Libor; Jurčáková, Klára; Kellnerová, Radka; Jaňour, Zbyněk

    2017-08-01

    This paper presents an extension of our previous wind-tunnel study (Nosek et al., 2016) in which we highlighted the need for investigation of the removal mechanisms of traffic pollution from all openings of a 3D street canyon. The extension represents the pollution flux (turbulent and advective) measurements at the lateral openings of three different 3D street canyons for the winds perpendicular and oblique to the along-canyon axis. The pollution was simulated by emitting a passive gas (ethane) from a homogeneous ground-level line source positioned along the centreline of the investigated street canyons. The street canyons were formed by courtyard-type buildings of two different regular urban-array models. The first model has a uniform building roof height, while the second model has a non-uniform roof height along each building's wall. The mean flow and concentration fields at the canyons' lateral openings confirm the findings of other studies that the buildings' roof-height variability at the intersections plays an important role in the dispersion of the traffic pollutants within the canyons. For the perpendicular wind, the non-uniform roof-height canyon appreciably removes or entrains the pollutant through its lateral openings, contrary to the uniform canyon, where the pollutant was removed primarily through the top. The analysis of the turbulent mass transport revealed that the coherent flow structures of the lateral momentum transport correlate with the ventilation processes at the lateral openings of all studied canyons. These flow structures coincide at the same areas and hence simultaneously transport the pollutant in opposite directions. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  3. Validation of CATHARE for gas-cooled reactors

    International Nuclear Information System (INIS)

    Fabrice Bentivoglio; Ola Widlund; Manuel Saez

    2005-01-01

    Full text of publication follows: Extensively validated and qualified for light-water reactor safety studies, the thermo-hydraulics code CATHARE has been adapted to deal also with gas-cooled reactor applications. In order to validate the code for these novel applications, CEA (Commissariat a l'Energie Atomique) has initiated an ambitious long-term experimental program. The foreseen experimental facilities range from small-scale loops for physical correlations, to component technology and system demonstration loops. In the short-term perspective, CATHARE is being validated against existing experimental data, in particular from the German power plant Oberhausen II and the South African Pebble-Bed Micro Model (PBMM). Oberhausen II, operated by the German utility EVO, is a 50 MW(e) direct-cycle Helium turbine plant. The power source is a gas burner rather than a nuclear reactor core, but the power conversion system resembles those of the GFR (Gas-cooled Fast Reactor) and other high-temperature reactor concepts. Oberhausen II was operated for more than 100 000 hours between 1974 and 1988. Design specifications, drawings and experimental data have been obtained through the European HTR project, offering a unique opportunity to validate CATHARE on a large-scale Brayton cycle. Available measurements of temperatures, pressures and mass flows throughout the circuit have allowed a very comprehensive thermohydraulic description of the plant, in steady-state conditions as well as during transients. The Pebble-Bed Micro Model (PBMM) is a small-scale model conceived to demonstrate the operability and control strategies of the South African PBMR concept. The model uses Nitrogen instead of Helium, and an electrical heater with a maximum rating of 420 kW. As the full-scale PBMR, the PBMM loop features three turbines and two compressors on the primary circuit, located on three separate shafts. The generator, however, is modelled by a third compressor on a separate circuit, with a

  4. Gas-cooled reactor application for a university campus

    International Nuclear Information System (INIS)

    Colak, Ue.; Kadiroghlu, O.K.; Soekmen, C.N.; Schmitt, H.

    1991-01-01

    Large urban areas with unfavourable topographic and meteorological conditions suffer severe air pollution during the winter months. Use of low grade lignites, imported higher quality coal or imported fuel oil are the sources of air pollution in the form of sulphur dioxide, fly ash and soot. Large housing complexes or old and historical locations within the city are in need of pollution free centralized district heating systems. Natural gas imported from the Soviet Union is a solution for this problem. Lack of gas distribution network for high pressure gas within the city is the main bottle-neck for the heating systems utilizing natural gas. Concern of the safety of flammable high pressure gas circulating within the city is another drawback for the natural gas heating systems. Nuclear district heating is an environmentally viable option worth looking into it. Localized urban nuclear heating is an interesting solution for large urban areas with old and historical character. The results of a feasibility study on the HGR application for the Hacettepe University presented here, summarizes the concept of gas-cooled heating reactors specially designed for urban centers. The inherently safe characteristics of the pebble bed heating reactor makes localized urban nuclear heating a viable alternative to other heat sources. An economical analysis of various heat sources with equal power levels is done for the Beytepe campus of Hacettepe University in Ankara. Under special boundary conditions, the price for heat generation can be much lower for nuclear heating with GHR 20 than for hard coal or fuel oil. It is also possible that if the price escalation rate for natural gas exceeds 3%, then nuclear heating with GHR can be more competitive. It is concluded that the nuclear heating of Beytepe campus with a GHR 20 is feasible and economical. (author) 3 figs., 5 refs

  5. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  6. A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

    Science.gov (United States)

    Syarip; Po, L. C. C.

    2018-05-01

    In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

  7. Development of a Computer Code, PZRTR rev 1, for the Thermal Hydraulic Analysis of a Multi-Cavity Cold Gas Pressurizer for an Integral Reactor, SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jae Kwang; Kang, H. O.; Yoon, J.; Kim, K. K

    2006-12-15

    The concept of a Multi-cavity Cold Gas PressuriZeR(MCGPZR) is applied to the SMART: The pressurizer system includes in-vessel cavities and out-of-vessel gas cylinders holding the gas supply/vent system. The gas cylinders are connected to the one of the in-vessel cavities via piping with valves. A pressurizer is maintained at a cold temperature of less than about 120 .deg. C which is realized with coolers installed in and with wet thermal insulators installed on one of the cavities located inside the hot reactor vessel, to minimize the contribution of a steam partial pressure and is filled with nitrogen gas as a pressure-absorbing medium. The working medium and working temperature of the MCGPZR is totally different from that of a hot steam pressurizer of the commercial PWR. In addition, the MCGPZR is intended to be designed to meet a pressure transient during normal power operation (by its gas volume capacity) without using an active control system and during plant heatup/cooldown operation by using an active gas control (filling/venting) system. Therefore in order to evaluate the feasibility of the concept of the MCGPZR and its intended design goal, the thermal hydraulic behaviors and controllability of the MCGPZR during transients especially a heatup/cooldown operation must be analyzed. In this study, a thermal hydraulic transient analysis computer code, PZRTR rev 1, for the Reactor Coolant System(RCS) of an integral reactor composed of the MCGPZR, modular Once-Through Steam Generators(OTSGs), a core and a reactor coolant loop is developed. The pressurizer module (MCGPZR module) of the PZRTR rev 1 code is based on a two-fluid, nonhomogeneous, nonequilibrium model for the two-phase system behavior and the OTSG module is based on a homogeneous equilibrium model of the two-phase flow process. The core module is simply based on the axial power distributions and the reactor coolant loop is based on the temperature distributions. The code is currently dedicated for the

  8. Non-equilibrium plasma reactor for natrual gas processing

    International Nuclear Information System (INIS)

    Shair, F.H.; Ravimohan, A.L.

    1974-01-01

    A non-equilibrium plasma reactor for natural gas processing into ethane and ethylene comprising means of producing a non-equilibrium chemical plasma wherein selective conversion of the methane in natural gas to desired products of ethane and ethylene at a pre-determined ethane/ethylene ratio in the chemical process may be intimately controlled and optimized at a high electrical power efficiency rate by mixing with a recycling gas inert to the chemical process such as argon, helium, or hydrogen, reducing the residence time of the methane in the chemical plasma, selecting the gas pressure in the chemical plasma from a wide range of pressures, and utilizing pulsed electrical discharge producing the chemical plasma. (author)

  9. Nuclear power developments in the Asia-Pacific region

    International Nuclear Information System (INIS)

    Irwin, T.

    2001-01-01

    There are 438 nuclear power reactors operating in the world. Of these, 95 are in the Asia-Pacific region. Of the 36 reactors currently under construction in the world, 19 are in the Asia-Pacific region. Of the 44 planned reactors in the world, 36 are in this region. At the start of the 'New Nuclear Century' the Asia-Pacific region has become the main area for growth and innovation in nuclear power. This paper describes the nuclear power developments in each country and examines the status of the construction programme and the planned projects. Countries included are China, India, Japan, Democratic People's Republic of Korea (DPRK), Republic of Korea, Pakistan and Taiwan. New projects include the HTR in China, Advance Breeder Water Reactors in Japan, KEDO in the DPRK and the Advance Pressurised Water Reactor in the Republic of Korea

  10. Utility industry evaluation of the Modular High-Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; Bitel, J.S.; Tramm, T.R.; High, M.D.; Neils, G.H.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the Modular High Temperature Gas-Cooled Reactor plant design, a current design created by an industrial team led by General Atomics under Department of Energy sponsorship and with support provided by utilities through Gas-Cooled Reactor Associates. The utility industry team concluded that the plant design should be considered a viable application of an advanced nuclear concept and deserves continuing development. Specific comments and recommendations are provided as a contribution toward improving a very promising plant design. 2 refs

  11. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  12. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1.11, September 1981

    International Nuclear Information System (INIS)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts

  13. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  14. Experimental measurements in the BYU controlled profile reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tree, D.R.; Black, D.l.; Rigby, J.R.; McQuay, M.Q.; Webb, B.W. [Brigham Young University, Provo, UT (United States). Dept. of Mechanical Engineering

    1998-09-01

    Over the past decade the Controlled Profile Reactor (CPR) has been used to obtain extensive combustion data sets. CPR is a small scale (0.2-0.4 MW) combustion facility that has been used to obtain data for model validation, the testing of new combustion concepts, and the development of new combustion instruments. This review of the past ten years of research completed in the CPR includes a description of the reactor and instrumentation used, a summary of three experimental data sets which have been obtained in the reactor, and a description of novel tests and instrumentation. Measurements obtained include gas species, gas temperature, particle velocity, particle size, particle number density, particle-cloud temperature profiles, radiation and total heat flux to the wall, and wall temperatures. Species data include the measurement of CO, CO{sub 2}, NO, NO{sub x}, O{sub 2}, NH{sub 3} and HCN. The three combustion studies included one with natural gas combustion in a swirling flow, and two pulverized-coal combustion studies involving Utah Blind Canyon and Pittsburgh No. 8 coals. Most, but not all of the above measurements were obtained in each study. The second coal study involving the Pittsburgh No. 8 coal contained the most complete set of data and is described in detail. Novel combustion instrumentation includes the use of Coherent Anti-Stokes Raman Spectroscopy (CARS) to measure gas temperature. Novel combustion experiments include the measurement of NO{sub x} and burnout with coal-char blends. The measurements have led to an improved understanding of the combustion process and an understanding of the strengths and weaknesses associated with different aspects of comprehensive combustion models. 67 refs., 26 figs., 9 tabs.

  15. Gas release from a failed fuel pin after reactor shut-down

    International Nuclear Information System (INIS)

    Pshenichnikov, B.V.

    1975-01-01

    A mathematical model of gassing from a hypothetical core fuel element in the active zone of a stopped water-moderated reactor was analysed to investigate the process of liberation of gaseous fission products from an unpressurized fuel element. A one-dimensional problem was obtained as a result of the accepted hypotheses. A fault was assumed to have occured during reactor operation; at the same time, a vapour-gas mixture was considered to be present under the envelope at reactor working pressure by the moment of stoppage. An approximative estimation was made of the retardation time of pressure balancing at the open end of the fuel element, and also of the amount of total gas remaining in the gap under the fuel element envelope after pressure drop in the reactor. Estimation of retardation time permitted to conclude that pressure in the nonhermetic fuel element envelope follows pressure fluctuation in the reactor in the course of cooling, the retardation time of pressure balancing outside and inside the fuel element lasting but a few seconds

  16. A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1982-01-01

    Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

  17. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  18. Let's Bet on Sediments! Hudson Canyon Cruise--Grades 9-12. Focus: Sediments of Hudson Canyon.

    Science.gov (United States)

    National Oceanic and Atmospheric Administration (DOC), Rockville, MD.

    These activities are designed to teach about the sediments of Hudson Canyon. Students investigate and analyze the patterns of sedimentation in the Hudson Canyon, observe how heavier particles sink faster than finer particles, and learn that submarine landslides are avalanches of sediment in deep ocean canyons. The activity provides learning…

  19. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

  20. Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

    2005-01-01

    Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

  1. Heat transport and afterheat removal for gas cooled reactors under accident conditions

    International Nuclear Information System (INIS)

    2001-01-01

    The Co-ordinated Research Project (CRP) on Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions was organized within the framework of the International Working Group on Gas Cooled Reactors (IWGGCR). This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs) and supports the conduct of these activities. Advanced GCR designs currently being developed are predicted to achieve a high degree of safety through reliance on inherent safety features. Such design features should permit the technical demonstration of exceptional public protection with significantly reduced emergency planning requirements. For advanced GCRs, this predicted high degree of safety largely derives from the ability of the ceramic coated fuel particles to retain the fission products under normal and accident conditions, the safe neutron physics behaviour of the core, the chemical stability of the core and the ability of the design to dissipate decay heat by natural heat transport mechanisms without reaching excessive temperatures. Prior to licensing and commercial deployment of advanced GCRs, these features must first be demonstrated under experimental conditions representing realistic reactor conditions, and the methods used to predict the performance of the fuel and reactor must be validated against these experimental data. Within this CRP, the participants addressed the inherent mechanisms for removal of decay heat from GCRs under accident conditions. The objective of this CRP was to establish sufficient experimental data at realistic conditions and validated analytical tools to confirm the predicted safe thermal response of advance gas cooled reactors during accidents. The scope includes experimental and analytical investigations of heat transport by natural convection conduction and thermal

  2. Characteristics of flow and reactive pollutant dispersion in urban street canyons

    Science.gov (United States)

    Park, Soo-Jin; Kim, Jae-Jin; Kim, Minjoong J.; Park, Rokjin J.; Cheong, Hyeong-Bin

    2015-05-01

    In this study, the effects of aspect ratio defined as the ratio of building height to street width on the dispersion of reactive pollutants in street canyons were investigated using a coupled CFD-chemistry model. Flow characteristics for different aspect ratios were analyzed first. For each aspect ratio, six emission scenarios with different VOC-NOX ratios were considered. One vortex was generated when the aspect ratio was less than 1.6 (shallow street canyon). When the aspect ratio was greater than 1.6 (deep street canyon), two vortices were formed in the street canyons. Comparing to previous studies on two-dimensional street canyons, the vortex center is slanted toward the upwind building and reverse and downward flows are dominant in street canyons. Near the street bottom, there is a marked difference in flow pattern between in shallow and deep street canyons. Near the street bottom, reverse and downward flows are dominant in shallow street canyon and flow convergence exists near the center of the deep street canyons, which induces a large difference in the NOX and O3 dispersion patterns in the street canyons. NOX concentrations are high near the street bottom and decreases with height. The O3 concentrations are low at high NO concentrations near the street bottom because of NO titration. At a low VOC-NOX ratio, the NO concentrations are sufficiently high to destroy large amount of O3 by titration, resulting in an O3 concentration in the street canyon much lower than the background concentration. At high VOC-NOX ratios, a small amount of O3 is destroyed by NO titration in the lower layer of the street canyons. However, in the upper layer, O3 is formed through the photolysis of NO2 by VOC degradation reactions. As the aspect ratio increases, NOX (O3) concentrations averaged over the street canyons decrease (increase) in the shallow street canyons. This is because outward flow becomes strong and NOX flux toward the outsides of the street canyons increases

  3. Grand Canyon Monitoring and Research Center

    Science.gov (United States)

    Hamill, John F.

    2009-01-01

    The Grand Canyon of the Colorado River, one of the world's most spectacular gorges, is a premier U.S. National Park and a World Heritage Site. The canyon supports a diverse array of distinctive plants and animals and contains cultural resources significant to the region's Native Americans. About 15 miles upstream of Grand Canyon National Park sits Glen Canyon Dam, completed in 1963, which created Lake Powell. The dam provides hydroelectric power for 200 wholesale customers in six western States, but it has also altered the Colorado River's flow, temperature, and sediment-carrying capacity. Over time this has resulted in beach erosion, invasion and expansion of nonnative species, and losses of native fish. Public concern about the effects of Glen Canyon Dam operations prompted the passage of the Grand Canyon Protection Act of 1992, which directs the Secretary of the Interior to operate the dam 'to protect, mitigate adverse impacts to, and improve values for which Grand Canyon National Park and Glen Canyon National Recreation Area were established...' This legislation also required the creation of a long-term monitoring and research program to provide information that could inform decisions related to dam operations and protection of downstream resources.

  4. Metaphysics methods development for high temperature gas cooled reactor analysis

    International Nuclear Information System (INIS)

    Seker, V.; Downar, T. J.

    2007-01-01

    Gas cooled reactors have been characterized as one of the most promising nuclear reactor concepts in the Generation-IV technology road map. Considerable research has been performed on the design and safety analysis of these reactors. However, the calculational tools being used to perform these analyses are not state-of-the-art and are not capable of performing detailed three-dimensional analyses. This paper presents the results of an effort to develop an improved thermal-hydraulic solver for the pebble bed type high temperature gas cooled reactors. The solution method is based on the porous medium approach and the momentum equation including the modified Ergun's resistance model for pebble bed is solved in three-dimensional geometry. The heat transfer in the pebble bed is modeled considering the local thermal non-equilibrium between the solid and gas, which results in two separate energy equations for each medium. The effective thermal conductivity of the pebble-bed can be calculated both from Zehner-Schluender and Robold correlations. Both the fluid flow and the heat transfer are modeled in three dimensional cylindrical coordinates and can be solved in steady-state and time dependent. The spatial discretization is performed using the finite volume method and the theta-method is used in the temporal discretization. A preliminary verification was performed by comparing the results with the experiments conducted at the SANA test facility. This facility is located at the Institute for Safety Research and Reactor Technology (ISR), Julich, Germany. Various experimental cases are modeled and good agreement in the gas and solid temperatures is observed. An on-going effort is to model the control rod ejection scenarios as described in the OECD/NEA/NSC PBMR-400 benchmark problem. In order to perform these analyses PARCS reactor simulator code will be coupled with the new thermal-hydraulic solver. Furthermore, some of the other anticipated accident scenarios in the benchmark

  5. Observations and Predictability of Gap Winds in the Salmon River Canyon of Central Idaho, USA

    Directory of Open Access Journals (Sweden)

    Natalie S. Wagenbrenner

    2018-01-01

    Full Text Available This work investigates gap winds in a steep, deep river canyon prone to wildland fire. The driving mechanisms and the potential for forecasting the gap winds are investigated. The onset and strength of the gap winds are found to be correlated to the formation of an along-gap pressure gradient linked to periodic development of a thermal trough in the Pacific Northwest, USA. Numerical simulations are performed using a reanalysis dataset to investigate the ability of numerical weather prediction (NWP to simulate the observed gap wind events, including the timing and flow characteristics within the canyon. The effects of model horizontal grid spacing and terrain representation are considered. The reanalysis simulations suggest that horizontal grid spacings used in operational NWP could be sufficient for simulating the gap flow events given the regional-scale depression in which the Salmon River Canyon is situated. The strength of the events, however, is under-predicted due, at least in part, to terrain smoothing in the model. Routine NWP, however, is found to have mixed results in terms of forecasting the gap wind events, primarily due to problems in simulating the regional sea level pressure system correctly.

  6. Impact of roof height non-uniformity on pollutant transport between a street canyon and intersections

    International Nuclear Information System (INIS)

    Nosek, Štěpán; Kukačka, Libor; Jurčáková, Klára; Kellnerová, Radka; Jaňour, Zbyněk

    2017-01-01

    This paper presents an extension of our previous wind-tunnel study (Nosek et al., 2016) in which we highlighted the need for investigation of the removal mechanisms of traffic pollution from all openings of a 3D street canyon. The extension represents the pollution flux (turbulent and advective) measurements at the lateral openings of three different 3D street canyons for the winds perpendicular and oblique to the along-canyon axis. The pollution was simulated by emitting a passive gas (ethane) from a homogeneous ground-level line source positioned along the centreline of the investigated street canyons. The street canyons were formed by courtyard-type buildings of two different regular urban-array models. The first model has a uniform building roof height, while the second model has a non-uniform roof height along each building's wall. The mean flow and concentration fields at the canyons' lateral openings confirm the findings of other studies that the buildings' roof-height variability at the intersections plays an important role in the dispersion of the traffic pollutants within the canyons. For the perpendicular wind, the non-uniform roof-height canyon appreciably removes or entrains the pollutant through its lateral openings, contrary to the uniform canyon, where the pollutant was removed primarily through the top. The analysis of the turbulent mass transport revealed that the coherent flow structures of the lateral momentum transport correlate with the ventilation processes at the lateral openings of all studied canyons. These flow structures coincide at the same areas and hence simultaneously transport the pollutant in opposite directions. - Highlights: • The pollutant transport strongly depends on the roof-height arrangement. • The non-uniform canyons also remove the pollutants through their lateral openings. • The higher the upstream wall, the more pollutant is removed through the top. • The lateral coherent structures correlate

  7. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

  8. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, October 1, 1982-March 3, 1983

    International Nuclear Information System (INIS)

    1983-06-01

    This report provides descriptions and results of the technical effort during the first half of FY 83 on the Gas-Cooled Thermal Reactor Program. The work on Integration and Management (WBS 01) includes the preparation of the Advanced Systems Concept Evaluation Plan and the Advanced Systems Technology Development Plan in addition to the program management activities. The Market Definition (WBS 03) efforts considered the application of the Modular Reactor System with reforming (MRS-R) to the production of methanol and ammonia and the refining of petroleum. Within the Plant Technology (WBS 13) task there were activities to develop anlytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. In addition to the work on the advanced HTGR for process heat users, new activities were initiated in support of the HTGR-SC/C Lead plant Protect (WBS 30 and 31). The Plant Simulation task (WBS 31) was initiated to develop a computer code for simulation of plant operation and for plant transient systems analysis. The efforts on the advanced HTGR systems was performed under the Modular Systems task (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors

  9. Behaviour of gas cooled reactor fuel under accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    The Specialists Meeting on Behaviour of Gas Cooled Reactor Fuel under Accident Conditions was convened by the International Atomic Energy Agency on the recommendation of the International Working Group on Gas Cooled Reactors. The purpose of the meeting was to provide an international forum for the review of the development status and for the discussion on the behaviour of gas cooled reactor fuel under accident conditions and to identify areas in which additional research and development are still needed and where international co-operation would be beneficial for all involved parties. The meeting was attended by 45 participants from France, Germany, Japan, Switzerland, the Union of Soviet Socialists Republics, the United Kingdom, the United States of America, CEC and the IAEA. The meeting was subdivided into five technical sessions: Summary of Current Research and Development Programmes for Fuel; Fuel Manufacture and Quality Control; Safety Requirements; Modelling of Fission Product Release - Part I and Part II; Irradiation Testing/Operational Experience with Fuel Elements; Behaviour at Depressurization, Core Heat-up, Power Transients; Water/Steam Ingress - Part I and Part II. 22 papers were presented. A separate abstract was prepared for each of these papers. At the end of the meeting a round table discussion was held on Directions for Future R and D Work and International Co-operation. Refs, figs and tabs

  10. Gas-solid hydroxyethylation of potato starch in a stirred vibrating fluidized bed reactor

    NARCIS (Netherlands)

    Kuipers, N.J M; Stamhuis, Eize; Beenackers, A.A C M

    A novel reactor for modifying cohesive C-powders such as in the gas-solid hydroxyethylation of semidry potato starch is characterized, the so-called stirred vibrating fluidized bed reactor. Good fluidization characteristics are obtained in this reactor for certain combinations of stirring and

  11. An analysis of operational experience during low power and shutdown and a plan for addressing human reliability assessment issues

    International Nuclear Information System (INIS)

    Barriere, M.; Luckas, W.; Whitehead, D.; Ramey-Smith, A.

    1994-06-01

    Recent nuclear power plant events (e.g. Chernobyl, Diablo Canyon, and Vogtle) and US Nuclear Regulatory Commission (NRC) reports (e.g. NUREG-1449) have led to concerns regarding human reliability during low power and shutdown (LP ampersand S) conditions and limitations of human reliability analysis (HRA) methodologies in adequately representing the LP ampersand S environment. As a result of these concerns, the NRC initiated two parallel research projects to assess the influence of LP ampersand S conditions on human reliability through an analysis of operational experience at pressurized water reactors (PWRs) an boiling water reactors (BWRs). These research projects, performed by Brookhaven National Laboratory for PWRS, and Sandia National Laboratories for BWRs, identified unique aspects of human performance during LP ampersand S conditions and provided a program plan for research and development necessary to improve existing HRA methodologies. This report documents the results of the analysis of LP ampersand S operating experience and describes the improved HRA program plan

  12. Observations of the behaviour of gas in the wake behind a corner blockage in fast breeder reactor subassembly geometry

    International Nuclear Information System (INIS)

    Fukuzawa, Y.

    1979-07-01

    Observations were made of gas behaviour in the wake behind a 21% corner blockage in the subassembly geometry of a liquid metal fast breeder reactor. The test section used represented one half of the reactor fuel subassembly, divided along the vertical plane of symmetry through the blockage. A glass wall occupied the position of this plane. Water was allowed to flow between glass rods simulating fuel pins, the velocity being changed from 1.2 to 4.5 m/s. Argon was injected into the wake or into the flow upstream of the blockage, the injection rate being changed from 1 to 230 Ncm 3 /s (standard temperature and pressure). From the present experiment, the following is evident: The gas is accumulated in the wake behind the blockage, forming a gas cavity. The flow patterns of the two-phase mixture in the wake are classified into three types, depending on the liquid velocity. In the lower velocity range, a gas cavity cannot be present at rest, rising up through the wake as a single bubble due to buoyancy. In the higher velocity range, the gas cavity is broken up by the liquid flow forces, only small gas bubbles circulating in the wake. In the velocity range in between, the gas cavity is present in the wake. The cavity size depends on the gas injection rate and on the liquid velocity. From the results, the possibility of fuel failure caused by fission gas release at a blockage in the fast breeder reactor can be considered to depend on the operating conditions of the reactor, specially on the coolant velocity. (orig.) [de

  13. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  14. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  15. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  16. Discussion on Design Transients of Pebble-bed High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Wang Yan; Li Fu; Zheng Yanhua

    2014-01-01

    In order to assure high quality for the components and their supports in the reactor coolant system, etc., some thermal-hydraulic transient conditions will be selected and researched for equipment design evaluation to satisfy the requirements ASME code, which are based on the conservative estimates of the magnitude and frequency of the temperature and pressure transients resulting from various operating conditions in the plant. In the mature design on pressurized water reactor, five conditions are considered. For the developing advanced pebble-bed high temperature gas-cooled reactor(HTGR), its design and operation has much difference with other reactors, so the transients of the pebble-bed high temperature gas-cooled reactor have distinctive characteristics. In this paper, the possible design transients of the pebble-bed HTGR will be discussed, and the frequency of design transients for equipment fatigue analysis and stress analysis due to cyclic stresses is also studied. The results will provide support for the design and construct of the pebble-bed HTGR. (author)

  17. Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

    International Nuclear Information System (INIS)

    Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

    2003-01-01

    The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

  18. Wintertime Boundary Layer Structure in the Grand Canyon.

    Science.gov (United States)

    Whiteman, C. David; Zhong, Shiyuan; Bian, Xindi

    1999-08-01

    Wintertime temperature profiles in the Grand Canyon exhibit a neutral to isothermal stratification during both daytime and nighttime, with only rare instances of actual temperature inversions. The canyon warms during daytime and cools during nighttime more or less uniformly through the canyon's entire depth. This weak stability and temperature structure evolution differ from other Rocky Mountain valleys, which develop strong nocturnal inversions and exhibit convective and stable boundary layers that grow upward from the valley floor. Mechanisms that may be responsible for the different behavior of the Grand Canyon are discussed, including the possibility that the canyon atmosphere is frequently mixed to near-neutral stratification when cold air drains into the top of the canyon from the nearby snow-covered Kaibab Plateau. Another feature of canyon temperature profiles is the sharp inversions that often form near the canyon rims. These are generally produced when warm air is advected over the canyon in advance of passing synoptic-scale ridges.Wintertime winds in the main canyon are not classical diurnal along-valley wind systems. Rather, they are driven along the canyon axis by the horizontal synoptic-scale pressure gradient that is superimposed along the canyon's axis by passing synoptic-scale weather disturbances. They may thus bring winds into the canyon from either end at any time of day.The implications of the observed canyon boundary layer structure for air pollution dispersion are discussed.

  19. Role of fission gas release in reactor licensing

    International Nuclear Information System (INIS)

    1975-11-01

    The release of fission gases from oxide pellets to the fuel rod internal voidage (gap) is reviewed with regard to the required safety analysis in reactor licensing. Significant analyzed effects are described, prominent gas release models are reviewed, and various methods used in the licensing process are summarized. The report thus serves as a guide to a large body of literature including company reports and government documents. A discussion of the state of the art of gas release analysis is presented

  20. Contemporary sediment-transport processes in submarine canyons.

    Science.gov (United States)

    Puig, Pere; Palanques, Albert; Martín, Jacobo

    2014-01-01

    Submarine canyons are morphological incisions into continental margins that act as major conduits of sediment from shallow- to deep-sea regions. However, the exact mechanisms involved in sediment transfer within submarine canyons are still a subject of investigation. Several studies have provided direct information about contemporary sedimentary processes in submarine canyons that suggests different modes of transport and various triggering mechanisms. Storm-induced turbidity currents and enhanced off-shelf advection, hyperpycnal flows and failures of recently deposited fluvial sediments, dense shelf-water cascading, canyon-flank failures, and trawling-induced resuspension largely dominate present-day sediment transfer through canyons. Additionally, internal waves periodically resuspend ephemeral deposits within canyons and contribute to dispersing particles or retaining and accumulating them in specific regions. These transport processes commonly deposit sediments in the upper- and middle-canyon reaches for decades or centuries before being completely or partially flushed farther down-canyon by large sediment failures.

  1. Evaluation of high temperature gas reactor for demanding cogeneration load follow

    International Nuclear Information System (INIS)

    Yan, Xing L.; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Hino, Ryutaro

    2012-01-01

    Modular nuclear reactor systems are being developed around the world for new missions among which is cogeneration for industries and remote areas. Like existing fossil energy counterpart in these markets, a nuclear plant would need to demonstrate the feasibility of load follow including (1) the reliability to generate power and heat simultaneously and alone and (2) the flexibility to vary cogeneration rates concurrent to demand changes. This article reports the results of JAEA's evaluation on the high temperature gas reactor (HTGR) to perform these duties. The evaluation results in a plant design based on the materials and design codes developed with JAEA's operating test reactor and from additional equipment validation programs. The 600 MWt-HTGR plant generates electricity efficiently by gas turbine and 900degC heat by a topping heater. The heater couples via a heat transport loop to industrial facility that consumes the high temperature heat to yield heat product such as hydrogen fuel, steel, or chemical. Original control methods are proposed to automate transition between the load duties. Equipment challenges are addressed for severe operation conditions. Performance limits of cogeneration load following are quantified from the plant system simulation to a range of bounding events including a loss of either load and a rapid peaking of electricity. (author)

  2. Reactor modeling and process analysis for partial oxidation of natural gas

    NARCIS (Netherlands)

    Albrecht, B.A.

    2004-01-01

    This thesis analyses a novel process of partial oxidation of natural gas and develops a numerical tool for the partial oxidation reactor modeling. The proposed process generates syngas in an integrated plant of a partial oxidation reactor, a syngas turbine and an air separation unit. This is called

  3. Simulation of Water Gas Shift Zeolite Membrane Reactor

    Science.gov (United States)

    Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

    2017-07-01

    The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

  4. Development of the design of the High Temperature Gas Cooled Reactor experiment

    International Nuclear Information System (INIS)

    Lockett, G.E.; Huddle, R.A.U.

    1960-01-01

    Early in 1956 a small team was formed at the Atomic Energy Research Establishment, Harwell, to investigate the possibilities of the High Temperature Gas Cooled (H.T.G.C.) Reactor System. Although the primary objective of this team was to carry out a feasibility study of the system as a whole, it soon became apparent that, in addition to design studies and economic surveys of power producing reactors, the most appropriate approach to such a novel system was to carry out a design study of a relatively small (10 to 20 M.W.) Reactor Experiment, together with the necessary research and development work associated with such a reactor. This work proceeded within the U.K.A.E.A. during the three following years, and it was felt that realistic design proposals could be put forward with sufficient confidence to justify the detailed design and construction of a 20 M.W. Reactor Experiment. In April 1959 responsibility for this Reactor Experiment was taken over by the O.E.E.C. High Temperature Gas Cooled Reactor Project, the DRAGON Project, at the Atomic Energy Establishment, Winfrith, Dorset. In this Paper the research, development, and design work is reviewed, and the proposals for the Reactor Experiment are summarised. (author)

  5. A Study of the Effects of Gas Well Compressor Noise on Breeding Bird Populations of the Rattlesnake Canyon Habitat Management Area, San Juan County, New Mexico

    Energy Technology Data Exchange (ETDEWEB)

    LaGory, K.E.; Chang, Young-Soo; Chun, K.C.; Reeves, T.; Liebich, R.; Smith, K.

    2001-06-04

    This report, conducted from May through July 2000, addressed the potential effect of compressor noise on breeding birds in gas-production areas administered by the FFO, specifically in the Rattlesnake Canyon Habitat Management Area northeast of Farmington, New Mexico. The study was designed to quantify and characterize noise output from these compressors and to determine if compressor noise affected bird populations in adjacent habitat during the breeding season.

  6. A flow reactor setup for photochemistry of biphasic gas/liquid reactions

    Directory of Open Access Journals (Sweden)

    Josef Schachtner

    2016-08-01

    Full Text Available A home-built microreactor system for light-mediated biphasic gas/liquid reactions was assembled from simple commercial components. This paper describes in full detail the nature and function of the required building elements, the assembly of parts, and the tuning and interdependencies of the most important reactor and reaction parameters. Unlike many commercial thin-film and microchannel reactors, the described set-up operates residence times of up to 30 min which cover the typical rates of many organic reactions. The tubular microreactor was successfully applied to the photooxygenation of hydrocarbons (Schenck ene reaction. Major emphasis was laid on the realization of a constant and highly reproducible gas/liquid slug flow and the effective illumination by an appropriate light source. The optimized set of conditions enabled the shortening of reaction times by more than 99% with equal chemoselectivities. The modular home-made flow reactor can serve as a prototype model for the continuous operation of various other reactions at light/liquid/gas interfaces in student, research, and industrial laboratories.

  7. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail

  8. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  9. Verification of maximum radial power peaking factor due to insertion of FPM-LEU target in the core of RSG-GAS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Setyawan, Daddy, E-mail: d.setyawan@bapeten.go.id [Center for Assessment of Regulatory System and Technology for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia); Rohman, Budi [Licensing Directorate for Nuclear Installations and Materials, Indonesian Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada No. 8 Jakarta 10120 (Indonesia)

    2014-09-30

    Verification of Maximum Radial Power Peaking Factor due to insertion of FPM-LEU target in the core of RSG-GAS Reactor. Radial Power Peaking Factor in RSG-GAS Reactor is a very important parameter for the safety of RSG-GAS reactor during operation. Data of radial power peaking factor due to the insertion of Fission Product Molybdenum with Low Enriched Uranium (FPM-LEU) was reported by PRSG to BAPETEN through the Safety Analysis Report RSG-GAS for FPM-LEU target irradiation. In order to support the evaluation of the Safety Analysis Report incorporated in the submission, the assessment unit of BAPETEN is carrying out independent assessment in order to verify safety related parameters in the SAR including neutronic aspect. The work includes verification to the maximum radial power peaking factor change due to the insertion of FPM-LEU target in RSG-GAS Reactor by computational method using MCNP5and ORIGEN2. From the results of calculations, the new maximum value of the radial power peaking factor due to the insertion of FPM-LEU target is 1.27. The results of calculations in this study showed a smaller value than 1.4 the limit allowed in the SAR.

  10. Partly standing internal tides in a dendritic submarine canyon observed by an ocean glider

    Science.gov (United States)

    Hall, Rob A.; Aslam, Tahmeena; Huvenne, Veerle A. I.

    2017-08-01

    An autonomous ocean glider is used to make the first direct measurements of internal tides within Whittard Canyon, a large, dendritic submarine canyon system that incises the Celtic Sea continental slope and a site of high benthic biodiversity. This is the first time a glider has been used for targeted observations of internal tides in a submarine canyon. Vertical isopycnal displacement observations at different stations fit a one-dimensional model of partly standing semidiurnal internal tides - comprised of a major, incident wave propagating up the canyon limbs and a minor wave reflected back down-canyon by steep, supercritical bathymetry near the canyon heads. The up-canyon internal tide energy flux in the primary study limb decreases from 9.2 to 2.0 kW m-1 over 28 km (a dissipation rate of 1 - 2.5 ×10-7 Wkg-1), comparable to elevated energy fluxes and internal tide driven mixing measured in other canyon systems. Within Whittard Canyon, enhanced mixing is inferred from collapsed temperature-salinity curves and weakened dissolved oxygen concentration gradients near the canyon heads. It has previously been hypothesised that internal tides impact benthic fauna through elevated near-bottom current velocities and particle resuspension. In support of this, we infer order 20 cm s-1 near-bottom current velocities in the canyon and observe high concentrations of suspended particulate matter. The glider observations are also used to estimate a 1 °C temperature range and 12 μmol kg-1 dissolved oxygen concentration range, experienced twice a day by organisms on the canyon walls, due to the presence of internal tides. This study highlights how a well-designed glider mission, incorporating a series of tide-resolving stations at key locations, can be used to understand internal tide dynamics in a region of complex topography, a sampling strategy that is applicable to continental shelves and slopes worldwide.

  11. Safety analysis of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Mitake, Susumu; Ezaki, Masahiro; Suzuki, Katsuo; Takaya, Junichi; Shimazu, Akira

    1976-02-01

    Safety features of the experimental multi-purpose high-temperature gas-cooled reactor being developed in JAERI were studied or the basis of its preliminary conceptual design of the reactor plant. Covered are control of the plant in transients, plant behaviour in accidents, and functions of engineered safeguards, and also dynamics of the uprant and frequencies of the accidents. These studies have shown, (i) the reactor plant can be operated both in plant slave to reactor and reactor slave to plant control, (ii) stable control of

  12. The development of the gas cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Kernforschungszentrum Karlsruhe

    1975-01-01

    A survey of the present technological state is given on the basis of the developments made so far. Some milestones of development - e.g. the German gas breeder memorandum, the Gas Breeder Reactor Association the results of the BR-2 radiation experiments and of GfK-KWU design and safety studies - are described. The problems connected with a large store of plutonium are also discussed. (UA/AK) [de

  13. Simulation of a gas cooled reactor with the system code CATHARE

    International Nuclear Information System (INIS)

    Bentivoglio, Fabrice; Ruby, Alain; Geffraye, Genevieve; Messie, Anne; Saez, Manuel; Tauveron, Nicolas; Widlund, Ola

    2006-01-01

    In recent years the CEA has commissioned a wide range of feasibility studies of future advanced nuclear reactors, in particular gas-cooled reactors (GCR). This paper presents an overview of the use of the thermohydraulics code CATHARE in these activities. Extensively validated and qualified for pressurized water reactors, CATHARE has been adapted to deal also with gas-cooled reactor applications. Rather than branching off a separate GCR version of CATHARE, new features have been integrated as independent options in the standard version of the code, respecting the same stringent procedures for documentation and maintenance. CATHARE has evolved into an efficient tool for GCR applications, with first results in good agreement with existing experimental data and other codes. The paper give an example among the studies already carried out with CATHARE with the case of the Very High Temperature Reactor (VHTR) concepts. Current and future activities for experimental validation of CATHARE for GCR applications are also discussed. Short-term validation activities are also included with the assessment of the German utility Oberhausen II. For the long term, CEA has initiated an ambitious experimental program ranging from small scale loops for physical correlations to component technology and system demonstration loops. (authors)

  14. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

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

    International Nuclear Information System (INIS)

    1997-11-01

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

  16. Contribution to the modelling of gas-solid reactions and reactors

    International Nuclear Information System (INIS)

    Patisson, F.

    2005-09-01

    Gas-solid reactions control a great number of major industrial processes involving matter transformation. This dissertation aims at showing that mathematical modelling is a useful tool for both understanding phenomena and optimising processes. First, the physical processes associated with a gas-solid reaction are presented in detail for a single particle, together with the corresponding available kinetic grain models. A second part is devoted to the modelling of multiparticle reactors. Different approaches, notably for coupling grain models and reactor models, are illustrated through various case studies: coal pyrolysis in a rotary kiln, production of uranium tetrafluoride in a moving bed furnace, on-grate incineration of municipal solid wastes, thermogravimetric apparatus, nuclear fuel making, steel-making electric arc furnace. (author)

  17. The genetic source and timing of hydrocarbon formation in gas hydrate reservoirs in Green Canyon, Block GC955

    Science.gov (United States)

    Moore, M. T.; Darrah, T.; Cook, A.; Sawyer, D.; Phillips, S.; Whyte, C. J.; Lary, B. A.

    2017-12-01

    Although large volumes of gas hydrates are known to exist along continental slopes and below permafrost, their role in the energy sector and the global carbon cycle remains uncertain. Investigations regarding the genetic source(s) (i.e., biogenic, thermogenic, mixed sources of hydrocarbon gases), the location of hydrocarbon generation, (whether hydrocarbons formed within the current reservoir formations or underwent migration), rates of clathrate formation, and the timing of natural gas formation/accumulation within clathrates are vital to evaluate economic potential and enhance our understanding of geologic processes. Previous studies addressed some of these questions through analysis of conventional hydrocarbon molecular (C1/C2+) and stable isotopic (e.g., δ13C-CH4, δ2H-CH4, δ13C-CO2) composition of gases, water chemistry and isotopes (e.g., major and trace elements, δ2H-H2O, δ18O-H2O), and dissolved inorganic carbon (δ13C-DIC) of natural gas hydrate systems to determine proportions of biogenic and thermogenic gas. However, the effects from contributions of mixing, transport/migration, methanogenesis, and oxidation in the subsurface can complicate the first-order application of these techniques. Because the original noble gas composition of a fluid is preserved independent of microbial activity, chemical reactions, or changes in oxygen fugacity, the integration of noble gas data can provide both a geochemical fingerprint for sources of fluids and an additional insight as to the uncertainty between effects of mixing versus post-genetic modification. Here, we integrate inert noble gases (He, Ne, Ar, and associated isotopes) with these conventional approaches to better constrain the source of gas hydrate formation and the residence time of fluids (porewaters and natural gases) using radiogenic 4He ingrowth techniques in cores from two boreholes collected as part of the University of Texas led UT-GOM2-01 drilling project. Pressurized cores were extracted from

  18. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  19. Street canyon aerosol pollutant transport measurements.

    Science.gov (United States)

    Longley, I D; Gallagher, M W; Dorsey, J R; Flynn, M; Bower, K N; Allan, J D

    2004-12-01

    Current understanding of dispersion in street canyons is largely derived from relatively simple dispersion models. Such models are increasingly used in planning and regulation capacities but are based upon a limited understanding of the transport of substances within a real canyon. In recent years, some efforts have been made to numerically model localised flow in idealised canyons (e.g., J. Appl. Meteorol. 38 (1999) 1576-89) and stepped canyons (Assimakopoulos V. Numerical modelling of dispersion of atmospheric pollution in and above urban canopies. PhD thesis, Imperial College, London, 2001) but field studies in real canyons are rare. To further such an understanding, a measurement campaign has been conducted in an asymmetric street canyon with busy one-way traffic in central Manchester in northern England. The eddy correlation method was used to determine fluxes of size-segregated accumulation mode aerosol. Measurements of aerosol at a static location were made concurrently with measurements on a platform lift giving vertical profiles. Size-segregated measurements of ultrafine and coarse particle concentrations were also made simultaneously at various heights. In addition, a small mobile system was used to make measurements of turbulence at various pavement locations within the canyon. From this data, various features of turbulent transport and dispersion in the canyon will be presented. The concentration and the ventilation fluxes of vehicle-related aerosol pollutants from the canyon will be related to controlling factors. The results will also be compared with citywide ventilation data from a separate measurement campaign conducted above the urban canopy.

  20. Geohydrology of White Rock Canyon of the Rio Grande from Otowi to Frijoles Canyon

    International Nuclear Information System (INIS)

    Purtymun, W.D.; Peters, R.J.; Owens, J.W.

    1980-12-01

    Twenty-seven springs discharge from the Totavi Lentil and Tesuque Formation in White Rock Canyon. Water generally acquires its chemical characteristics from rock units that comprise the spring aquifer. Twenty-two of the springs are separated into three groups of similar aquifer-related chemical quality. The five remaining springs make up a fourth group with a chemical quality that differs due to localized conditions in the aquifer. Localized conditions may be related to recharge or discharge in or near basalt intrusion or through faults. Streams from Pajarito, Ancho, and Frijoles Canyons discharge into the Rio Grande in White Rock Canyon. The base flow in the streams is from springs. Sanitary effluent in Mortandad Canyon from the treatment plant at White Rock also reaches the Rio Grande

  1. 78 FR 48670 - Boulder Canyon Project

    Science.gov (United States)

    2013-08-09

    ... DEPARTMENT OF ENERGY Western Area Power Administration Boulder Canyon Project AGENCY: Western Area... Canyon Project (BCP) electric service provided by the Western Area Power Administration (Western). The... INFORMATION: Hoover Dam, authorized by the Boulder Canyon Project Act (45 Stat. 1057, December 21, 1928), sits...

  2. 77 FR 48151 - Boulder Canyon Project

    Science.gov (United States)

    2012-08-13

    ... DEPARTMENT OF ENERGY Western Area Power Administration Boulder Canyon Project AGENCY: Western Area... Canyon Project (BCP) electric service provided by the Western Area Power Administration (Western). The... INFORMATION: Hoover Dam, authorized by the Boulder Canyon Project Act (45 Stat. 1057, December 21, 1928), sits...

  3. Features, present condition of development and future scope on the high temperature gas reactor as an innovative one

    International Nuclear Information System (INIS)

    Shiozawa, Shusaku

    2001-01-01

    The high temperature gas reactor has some features without previous reactors such as high temperature capable of taking-out, high specific safety, feasibility adaptable to versatile fuel cycle, and so on. Then, it is expected to be an innovative reactor to contribute to diversification of energy supply and expansion of energy application field. In Japan, under the HTTR (high temperature engineering test reactor) plan, construction of HTTR, which is the first high temperature gas reactor in Japan, was finished and its output upgrading test has been promoted. And, on the HTTR plan, together with promotion of full power operation, reactor performance tests, safety proof test, and so on, it is planned to carry out study on application of the high temperature heat such as hydrogen production and so on to aim to practise establishment and upgrading of technologies on high temperature gas reactor in Japan. Here were introduced features and present condition of development of the high temperature gas reactor as an innovative type reactor and described role and future scope in Japan. (G.K.)

  4. Hydrogen enrichment and separation from synthesis gas by the use of a membrane reactor

    International Nuclear Information System (INIS)

    Sanchez, J.M.; Barreiro, M.M.; Marono, M.

    2011-01-01

    One of the objectives of the CHRISGAS project was to study innovative gas separation and gas upgrading systems that have not been developed sufficiently yet to be tested at a demonstration scale within the time frame of the project, but which show some attractive merits and features for further development. In this framework CIEMAT studied, at bench scale, hydrogen enrichment and separation from syngas by the use of membranes and membrane catalytic reactors. In this paper results about hydrogen separation from synthesis gas by means of selective membranes are presented. Studies dealt with the evaluation of permeation and selectivity to hydrogen of prepared and pre-commercial Pd-based membranes. Whereas prepared membranes turned out to be non-selective, due to discontinuities of the palladium layer, studies conducted with the pre-commercial membrane showed that by means of a membrane reactor it is possible to completely separate hydrogen from the other gas components and produce pure hydrogen as a permeate stream, even in the case of complex reaction system (H 2 /CO/CO 2 /H 2 O) under WGS conditions gas mixtures. The advantages of using a water-gas shift membrane reactor (MR) over a traditional fixed bed reactor (TR) have also been studied. The experimental device included the pre-commercial Pd-based membrane and a commercial high temperature Fe-Cr-based, WGS catalyst, which was packed in the annulus between the membrane and the reactor outer shell. Results show that in the MR concept, removal of H 2 from the reaction side has a positive effect on WGS reaction, reaching higher CO conversion than in a traditional packed bed reactor at a given temperature. On increasing pressure on the reaction side permeation is enhanced and hence carbon monoxide conversion increases. -- Highlights: → H 2 enrichment and separation using a bench-scale membrane reactor MR is studied. → Permeation and selectivity to H 2 of Pd-based membranes was determined. → Complete separation

  5. High-temperature Gas Reactor (HTGR)

    Science.gov (United States)

    Abedi, Sajad

    2011-05-01

    General Atomics (GA) has over 35 years experience in prismatic block High-temperature Gas Reactor (HTGR) technology design. During this period, the design has recently involved into a modular have been performed to demonstrate its versatility. This versatility is directly related to refractory TRISO coated - particle fuel that can contain any type of fuel. This paper summarized GA's fuel cycle studies individually and compares each based upon its cycle sustainability, proliferation-resistance capabilities, and other performance data against pressurized water reactor (PWR) fuel cycle data. Fuel cycle studies LEU-NV;commercial HEU-Th;commercial LEU-Th;weapons-grade plutonium consumption; and burning of LWR waste including plutonium and minor actinides in the MHR. results show that all commercial MHR options, with the exception of HEU-TH, are more sustainable than a PWR fuel cycle. With LEU-NV being the most sustainable commercial options. In addition, all commercial MHR options out perform the PWR with regards to its proliferation-resistance, with thorium fuel cycle having the best proliferation-resistance characteristics.

  6. IAEA activities in gas-cooled reactor technology development

    International Nuclear Information System (INIS)

    Cleveland, J.; Kupitz, J.

    1992-01-01

    The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

  7. Heterogeneous gas core reactor

    International Nuclear Information System (INIS)

    Diaz, N.J.; Dugan, E.T.

    1983-01-01

    A heterogeneous gas core nuclear reactor is disclosed comprising a core barrel provided interiorly with an array of moderator-containing tubes and being otherwise filled with a fissile and/or fertile gaseous fuel medium. The fuel medium may be flowed through the chamber and through an external circuit in which heat is extracted. The moderator may be a fluid which is flowed through the tubes and through an external circuit in which heat is extracted. The moderator may be a solid which may be cooled by a fluid flowing within the tubes and through an external heat extraction circuit. The core barrel is surrounded by moderator/coolant material. Fissionable blanket material may be disposed inwardly or outwardly of the core barrel

  8. Parametric Investigation of Brayton Cycle for High Temperature Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Chang Oh

    2004-01-01

    The Idaho National Engineering and Environmental Laboratory (INEEL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. In this project, we are investigating helium Brayton cycles for the secondary side of an indirect energy conversion system. Ultimately we will investigate the improvement of the Brayton cycle using other fluids, such as supercritical carbon dioxide. Prior to the cycle improvement study, we established a number of baseline cases for the helium indirect Brayton cycle. These cases look at both single-shaft and multiple-shaft turbomachinery. The baseline cases are based on a 250 MW thermal pebble bed HTGR. The results from this study are applicable to other reactor concepts such as a very high temperature gas-cooled reactor (VHTR), fast gas-cooled reactor (FGR), supercritical water reactor (SWR), and others. In this study, we are using the HYSYS computer code for optimization of the helium Brayton cycle. Besides the HYSYS process optimization, we performed parametric study to see the effect of important parameters on the cycle efficiency. For these parametric calculations, we use a cycle efficiency model that was developed based on the Visual Basic computer language. As a part of this study we are currently investigated single-shaft vs. multiple shaft arrangement for cycle efficiency and comparison, which will be published in the next paper. The ultimate goal of this study is to use supercritical carbon dioxide for the HTGR power conversion loop in order to improve the cycle efficiency to values great than that of the helium Brayton cycle. This paper includes preliminary calculations of the steady state overall Brayton cycle efficiency based on the pebble bed reactor reference design (helium used as the working fluid) and compares those results with an initial calculation of a CO2 Brayton cycle

  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. Saclay Reactor: acquired knowledge by two years experience in heat transfer using compressed gas

    International Nuclear Information System (INIS)

    Yvon, J.

    1955-01-01

    Describes the conception and functioning of a new reactor (EL-2) using compressed gas as primary coolant. The aim of the use of compressed gas as primary coolant is to reduce the quantity of heavy water used in the functioning of the reactor. Description of the reactor vessel (dimensions, materials, reflector and protection). Description of the cells and the circulation of the gas within the cells. A complete explanation of the control and regulating of the reaction by the ionization chamber is given. Heavy water is used as modulator: it describes the heavy water system and its recombination system. The fuel slugs are cooled by compressed gas: its system is described as well as the blower and the heat exchanger system. Water is supplied by a cooling tower which means the reactor power is dependant of the atmospheric conditions. Particular attention has been given to the tightness of the different systems used. The relation between neutron flow and the thermal output is discussed: the thermal output can be calculated by measuring the gas flow and its heating or by measuring the neutron flow within the reactor, both methods gives closed results. Reactivity study: determination of the different factors which induce a variation of reactivity. Heat transfer: discussion on the use of different heat transfer systems, determination of the required chemical and physical properties of the primary coolant as well as the discussion of the nuclear and thermal requirements for the choice of it. A comparison between the use of nitrogen and carbon dioxide gas shows an advantage in using nitrogen with the existing knowledge. Reflexion on the relevance of this work and the future perspectives of the use of compressed gas as primary coolant. (M.P.)

  11. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    Locke, B

    1998-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  12. Reticulated Vitreous Carbon Electrodes for Gas Phase Pulsed Corona Reactors

    National Research Council Canada - National Science Library

    LOCKE, B

    1999-01-01

    A new design for gas phase pulsed corona reactors incorporating reticulated vitreous carbon electrodes is demonstrated to be effective for the removal of nitrogen oxides from synthetic air mixtures...

  13. Durable terrestrial bedrock predicts submarine canyon formation

    Science.gov (United States)

    Smith, Elliot; Finnegan, Noah J.; Mueller, Erich R.; Best, Rebecca J.

    2017-01-01

    Though submarine canyons are first-order topographic features of Earth, the processes responsible for their occurrence remain poorly understood. Potentially analogous studies of terrestrial rivers show that the flux and caliber of transported bedload are significant controls on bedrock incision. Here we hypothesize that coarse sediment load could exert a similar role in the formation of submarine canyons. We conducted a comprehensive empirical analysis of canyon occurrence along the West Coast of the contiguous United States which indicates that submarine canyon occurrence is best predicted by the occurrence of durable crystalline bedrock in adjacent terrestrial catchments. Canyon occurrence is also predicted by the flux of bed sediment to shore from terrestrial streams. Surprisingly, no significant correlation was observed between canyon occurrence and the slope or width of the continental shelf. These findings suggest that canyon incision is promoted by greater yields of durable terrestrial clasts to the shore.

  14. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Science.gov (United States)

    Grodzki, Marcin; Darnowski, Piotr; Niewiński, Grzegorz

    2017-12-01

    The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an `early design' variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit). A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  15. Monte Carlo Analysis of the Battery-Type High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Grodzki Marcin

    2017-12-01

    Full Text Available The paper presents a neutronic analysis of the battery-type 20 MWth high-temperature gas cooled reactor. The developed reactor model is based on the publicly available data being an ‘early design’ variant of the U-battery. The investigated core is a battery type small modular reactor, graphite moderated, uranium fueled, prismatic, helium cooled high-temperature gas cooled reactor with graphite reflector. The two core alternative designs were investigated. The first has a central reflector and 30×4 prismatic fuel blocks and the second has no central reflector and 37×4 blocks. The SERPENT Monte Carlo reactor physics computer code, with ENDF and JEFF nuclear data libraries, was applied. Several nuclear design static criticality calculations were performed and compared with available reference results. The analysis covered the single assembly models and full core simulations for two geometry models: homogenous and heterogenous (explicit. A sensitivity analysis of the reflector graphite density was performed. An acceptable agreement between calculations and reference design was obtained. All calculations were performed for the fresh core state.

  16. The Asia Pacific gas market: a question of balance

    International Nuclear Information System (INIS)

    Russell Jacobs

    1997-01-01

    The underlying need for additional supplies of natural gas, both and pipeline and LNG , will be continue to expand in the Asia Pacific region. Spurred by expected development of LNG markets in Thailand,India,and coastal China, the demand for LNG could more than double by 2010. To meet the LNG needs of the future, numerous LNG grass roots and expansion projects are underway or firmly planned. Collectively , these projects could supply nearly million tonnes of additional LNG by 2005-2010. If new geographical markets can not be developed (for whatever reasons) during this time frame, however, some currently planned projects could falter or be under utilized. (Author) 3 figs

  17. The Asia Pacific gas market: a question of balance

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, Russell [Purvin and Gertz, Inc., Dallas, TX (United States)

    1997-06-01

    The underlying need for additional supplies of natural gas, both and pipeline and LNG , will be continue to expand in the Asia Pacific region. Spurred by expected development of LNG markets in Thailand,India,and coastal China, the demand for LNG could more than double by 2010. To meet the LNG needs of the future, numerous LNG grass roots and expansion projects are underway or firmly planned. Collectively , these projects could supply nearly million tonnes of additional LNG by 2005-2010. If new geographical markets can not be developed (for whatever reasons) during this time frame, however, some currently planned projects could falter or be under utilized. (Author) 3 figs.

  18. Integrated gasification gas combined cycle plant with membrane reactors: Technological and economical analysis

    International Nuclear Information System (INIS)

    Amelio, Mario; Morrone, Pietropaolo; Gallucci, Fausto; Basile, Angelo

    2007-01-01

    In the present work, the capture and storage of carbon dioxide from the fossil fuel power plant have been considered. The main objective was to analyze the thermodynamic performances and the technological aspects of two integrated gasification gas combined cycle plants (IGCC), as well as to give a forecast of the investment costs for the plants and the resulting energy consumptions. The first plant considered is an IGCC* plant (integrated gasification gas combined cycle plant with traditional shift reactors) characterized by the traditional water gas shift reactors and a CO 2 physical adsorption system followed by the power section. The second one is an IGCC M plant (integrated gasification gas combined cycle plant with membrane reactor) where the coal thermal input is the same as the first one, but the traditional shift reactors and the physical adsorption unit are replaced by catalytic palladium membrane reactors (CMR). In the present work, a mono-dimensional computational model of the membrane reactor was proposed to simulate and evaluate the capability of the IGCC M plant to capture carbon dioxide. The energetic performances, efficiency and net power of the IGCC* and IGCC M plants were, thus, compared, assuming as standard a traditional IGCC plant without carbon dioxide capture. The economical aspects of the three plants were compared through an economical analysis. Since the IGCC* and IGCC M plants have additional costs related to the capture and disposal of the carbon dioxide, a Carbon Tax (adopted in some countries like Sweden) proportional to the number of kilograms of carbon dioxide released in the environment was assumed. According to the economical analysis, the IGCC M plant proved to be more convenient than the IGCC* one

  19. Proposal of a fast gas-cooled reactor using transuranics

    International Nuclear Information System (INIS)

    Macedo, Anderson Altair Pinheiro de

    2016-01-01

    In the last two decades, nations that have invested in research and energy generation through nuclear source have devoted part of their efforts in developing new technologies for nuclear reactors. Part of this investment focuses on new material testing, particularly regarding new fuels. In a world view that breaths sustainability, the reprocess and reuse of spent fuel from conventional reactors comes alive in nuclear technology, presenting itself as a real alternative of energy source for the latest generation of reactors. Different concepts of fourth generation reactors have been proposed and must meet some basic requirements, such as: extended burnup, improvement of passive safety, better radioactive waste management, possibility to use reprocessed fuel and proliferation resistance. In this context, the GFR (Gas-cooled Fast Reactor) is one of the future promises, presenting satisfactory neutronic results on the use of type of fuel (U, Pu) C. In the present work, the fuel of a traditional GFR reactor that uses (U, Pu)C was sub was replaced by a transuranic reprocessed fuel (TRU), obtained by non-proliferation reprocessing technology. The UO 2 fuel initially enriched by 3.1% was burned in a standard PWR, with full burn of 33,000 MWd/T. Afterward it was left in a pool for 5 years and finally reprocessed by UREX + method. Two fuels were studied and evaluated, one diluted with depleted uranium (U, TRU)C, and the other diluted in thorium (Th, TRU)C. Assessments were done in steady state and as well as during burning and were compared with results obtained using the standard fuel, (U, Pu) C. The outcome shows that the use of TRU as a fuel, in GFR type reactors, is a real possibility. The research was done using the SCALE 6.0 code modules. (author)

  20. Deformation near the Casa Diablo geothermal well field and related processes Long Valley caldera, Eastern California, 1993-2000

    Science.gov (United States)

    Howle, J.F.; Langbein, J.O.; Farrar, C.D.; Wilkinson, S.K.

    2003-01-01

    Regional first-order leveling lines, which extend from Lee Vining, CA, to Tom's Place, CA, have been surveyed periodically since 1957 by the U.S. Geological Survey (USGS), the National Geodetic Survey (NGS), and Caltrans. Two of the regional survey lines, or leveling networks, intersect at the Casa Diablo geothermal well field. These leveling networks, referenced to a distant bench mark (C916) near Lee Vining, provide time-series vertical control data of land-surface deformation that began around 1980. These data are also useful for delineating localized subsidence at Casa Diablo related to reservoir pressure and temperature changes owing to geothermal development that began in 1985. A comparison of differences in bench-mark elevations for five time periods between 1983 and 1997 shows the development and expansion of a subsidence bowl at Casa Diablo. The subsidence coincides spatially with the geothermal well field and temporally with the increased production rates and the deepening of injection wells in 1991, which resulted in an increase in the rate of pressure decline. The subsidence, superimposed on a broad area of uplift, totaled about 310 mm by 1997. The USGS established orthogonal tilt arrays in 1983 to better monitor deformation across the caldera. One tilt array (DBR) was established near what would later become the Casa Diablo geothermal well field. This array responded to magmatic intrusions prior to geothermal development, tilting away from the well field. With the start of geothermal fluid extraction in 1985, tilt at the DBR array reversed direction and began tilting into the well field. In 1991, geothermal power production was increased by a factor of four, and reservoir pressures began a period of steep decline. These changes caused a temporary three-fold increase in the tilt rate. The tilt rate became stable in 1993 and was about 40% lower than that measured in 1991-1992, but still greater than the rates measured during 1985-1990. Data from the

  1. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Marmier, Alain

    2012-01-01

    high temperature irradiation to high burn-ups with fission gas release measurements. To this end, the HFR-EU1 fuel irradiation in the High Flux Reactor (HFR) Petten (2006-2010) explored the potential for high performance and high burn-up of existing German fuel (3 pebbles produced for the AVR reactor at the German research centre Juelich) and newly produced Chinese fuel (2 pebbles produced by INET for use in the HTR-10 test reactor in China). These five pebbles were irradiated for 445 days in separately controlled capsules, while the fission gas release was monitored by gamma spectrometry thus enabling evaluation of the characteristic release over birth fraction, indicative for the health of the fuel. In none of the pebbles, abnormally increased fission gas release was observed indicating that all of the approx. 45,000 coated particles in the pebbles had remained intact. The results presented in this thesis cover the first 332 days of irradiation. While HFR-EU1 was dedicated to a particularly high burn-up, HFR-EU1bis, performed between 2004 and 2005, investigated extremely high temperature for steady-state conditions. The comparison of both experiments confirms that temperature plays a decisive part in fuel performance and integrity. The peak fuel temperature in pebbles can be lowered with the so-called w allpaper fuel , in which the coated fuel particles are arranged in a spherical shell within a pebble. This wallpaper concept also enhances neutronic performance through improved neutron economy, resulting in reduced fissile material and/or enrichment needs or providing the potential to achieve higher burn-up. To quantify these improvements, calculations were performed using the Monte Carlo neutron transport and depletion codes MCNP/MCB (to assess conversion ratio, temperature coefficient of reactivity and neutron multiplication) and PANTHERMIX (for fuel cycle in steady state conditions and loss of coolant accident calculations). Based on PANTHERMIX steady

  2. Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

    International Nuclear Information System (INIS)

    Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

    2007-01-01

    According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

  3. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1985-07-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity.

  4. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    1985-01-01

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  5. A charge regulating system for turbo-generator gas-cooled high-temperature reactor power stations

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1975-01-01

    The invention relates to a regulating system for gas-cooled high-temperature reactors power stations (helium coolant), equipped with several steam-boilers, each of which deriving heat from a corresponding cooling-gas flow circulating in the reactor, so as to feed superheated steam into a main common steam-manifold and re-superheated steam into a re-superheated hot common manifold [fr

  6. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  7. Enhancing the production of hydrogen via water-gas shift reaction using Pd-based membrane reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mendes, Diogo; Chibante, Vania; Mendes, Adelio; Madeira, Luis M. [LEPAE, Chemical Engineering Department, Faculty of Engineering, University of Porto, Rua Dr. Roberto Frias, 4200-465 Porto (Portugal); Zheng, Ju-Meng [Dutch Separation Technology Institute (DSTI), 3800 AE Amersfoort (Netherlands); Tosti, Silvano; Borgognoni, Fabio [ENEA, Unita Tecnica Fusione, C.R. ENEA Frascati, Via E. Fermi 45, Frascati (RM) I-00044 (Italy)

    2010-11-15

    In this work, it is described an experimental study regarding the performance of a Pd-Ag membrane reactor recently proposed and suitable for the production of ultra-pure hydrogen. A dense metallic permeator tube was assembled by an innovative annealing and diffusion welding technique from a commercial flat sheet membrane of Pd-Ag. A ''finger-like'' configuration of the self-supported membrane has been designed and used as a packed-bed membrane reactor (MR) for producing ultra-pure hydrogen via water-gas shift reaction (WGS). A CuO/ZnO/Al{sub 2}O{sub 3} catalyst, from REB Research and Consulting, was used for packing the WGS membrane reactor. The performance of the reactor was evaluated in terms of CO conversion and H{sub 2} recovery in a wide range of conditions: temperature from 200 C to 300 C, feed pressure from 1.0 bar to 4.0 bar, vacuum and sweep-gas modes and with a simulated reformate feed (4.70% CO, 34.78% H{sub 2}O, 28.70% H{sub 2}, 10.16% CO{sub 2} balanced in N{sub 2}). Also, the effect of the reactants feed composition was investigated and discussed. CO conversions remained in most conditions above the thermodynamic equilibrium based on feed conditions. In particular, it is worth mentioning that around 100% of CO conversion and almost complete H{sub 2} recovery was achieved when operating the MR at 300 C with a GSHV = 1200 L{sub N} kg{sub cat}{sup -1} h{sup -1}, P{sub feed} = 4 bar, P{sub perm} = 3 bar and using 1000 mL{sub N} min{sup -1} of sweep-gas. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-11-01

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

  9. Plasmachemical oxidation processes in a hybrid gas-liquid electrical discharge reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lukes, Petr; Locke, Bruce R [Department of Chemical and Biomedical Engineering, FAMU-FSU College of Engineering, Florida State University, 2525 Pottsdamer Street, Tallahassee, Florida (United States)

    2005-11-21

    Oxidation processes induced in water by pulsed electrical discharges generated simultaneously in the gas phase in close proximity to the water surface and directly in the liquid were investigated in a hybrid series gas-liquid electrical discharge reactor. The mechanism of phenol degradation was studied through its dependence on the gas phase and liquid phase compositions using pure argon and oxygen atmospheres above the liquid and different initial pH values in the aqueous solution. Phenol degradation was significantly enhanced in the hybrid-series reactor compared with the phenol removal by the single-liquid phase discharge reactor. Under an argon atmosphere the mechanism of phenol degradation was mainly caused by the electrophilic attack of OH{center_dot} radicals produced by the liquid phase discharge directly in water and OH{center_dot} radicals produced by the gas phase discharge at the gas-liquid interface. Under an oxygen atmosphere the formation of gaseous ozone dominated over the formation of OH{center_dot} radicals, and the contribution of the gas phase discharge in this case was determined mainly by the dissolution of gaseous ozone into the water and its subsequent interaction with phenol. At high pH phenol was degraded, in addition to the direct attack by ozone, also through indirect reactions of OH{center_dot} radicals formed via a peroxone process by the decomposition of dissolved ozone by hydrogen peroxide produced by the liquid phase discharge. Such a mechanism was proved by the detection of cis,cis-muconic acid and pH-dependent degradation of phenol, which resulted in significantly higher removal of phenol from alkaline solution observed under oxygen atmosphere than in argon.

  10. Plasmachemical oxidation processes in a hybrid gas-liquid electrical discharge reactor

    International Nuclear Information System (INIS)

    Lukes, Petr; Locke, Bruce R

    2005-01-01

    Oxidation processes induced in water by pulsed electrical discharges generated simultaneously in the gas phase in close proximity to the water surface and directly in the liquid were investigated in a hybrid series gas-liquid electrical discharge reactor. The mechanism of phenol degradation was studied through its dependence on the gas phase and liquid phase compositions using pure argon and oxygen atmospheres above the liquid and different initial pH values in the aqueous solution. Phenol degradation was significantly enhanced in the hybrid-series reactor compared with the phenol removal by the single-liquid phase discharge reactor. Under an argon atmosphere the mechanism of phenol degradation was mainly caused by the electrophilic attack of OH· radicals produced by the liquid phase discharge directly in water and OH· radicals produced by the gas phase discharge at the gas-liquid interface. Under an oxygen atmosphere the formation of gaseous ozone dominated over the formation of OH· radicals, and the contribution of the gas phase discharge in this case was determined mainly by the dissolution of gaseous ozone into the water and its subsequent interaction with phenol. At high pH phenol was degraded, in addition to the direct attack by ozone, also through indirect reactions of OH· radicals formed via a peroxone process by the decomposition of dissolved ozone by hydrogen peroxide produced by the liquid phase discharge. Such a mechanism was proved by the detection of cis,cis-muconic acid and pH-dependent degradation of phenol, which resulted in significantly higher removal of phenol from alkaline solution observed under oxygen atmosphere than in argon

  11. French gas cooled reactor experience with moisture ingress

    International Nuclear Information System (INIS)

    Bastien, D.; Brie, M.

    1995-01-01

    During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

  12. Liquid metal reactor cover gas purification and analysis in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Allen, K J [Argonne National Laboratory, EBR-II Division, Idaho Falls, ID (United States); Meadows, G E; Schuck, W J [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA.

  13. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1987-01-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  14. Gas dynamics models for an oscillating gaseous core fission reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C.; Dam, H. van; Hoogenboom, J.E. (Interuniversitair Reactor Inst., Delft (Netherlands))

    1991-01-01

    Two one-dimensional models are developed for the investigation of the gas dynamical behaviour of the fuel gas in a cylindrical gaseous core fission reactor. By numerical and analytical calculations, it is shown that, for the case where a direct energy extraction mechanism (such as magneto-hydrodynamics (MHD)) is not present, increasing density oscillations occur in the gas. Also an estimate is made of the attainable direct energy conversion efficiency, for the case where a direct energy extraction mechanism is present. (author).

  15. Results of stainless steel canister corrosion studies and environmental sample investigations

    Energy Technology Data Exchange (ETDEWEB)

    Bryan, Charles R. [Sandia National Laboratories, Albuquerque, NM (United States); Enos, David [Sandia National Laboratories, Albuquerque, NM (United States)

    2014-12-01

    This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of used nuclear fuel. The work involves both characterization of the potential physical and chemical environment on the surface of the storage canisters and how it might evolve through time, and testing to evaluate performance of the canister materials under anticipated storage conditions. To evaluate the potential environment on the surface of the canisters, SNL is working with the Electric Power Research Institute (EPRI) to collect and analyze dust samples from the surface of in-service SNF storage canisters. In FY 13, SNL analyzed samples from the Calvert Cliffs Independent Spent Fuel Storage Installation (ISFSI); here, results are presented for samples collected from two additional near-marine ISFSI sites, Hope Creek NJ, and Diablo Canyon CA. The Hope Creek site is located on the shores of the Delaware River within the tidal zone; the water is brackish and wave action is normally minor. The Diablo Canyon site is located on a rocky Pacific Ocean shoreline with breaking waves. Two types of samples were collected: SaltSmart™ samples, which leach the soluble salts from a known surface area of the canister, and dry pad samples, which collected a surface salt and dust using a swipe method with a mildly abrasive ScotchBrite™ pad. The dry samples were used to characterize the mineralogy and texture of the soluble and insoluble components in the dust via microanalytical techniques, including mapping X-ray Fluorescence spectroscopy and Scanning Electron Microscopy. For both Hope Creek and Diablo Canyon canisters, dust loadings were much higher on the flat upper surfaces of the canisters than on the vertical sides. Maximum dust sizes collected at both sites were slightly larger than 20 μm, but Phragmites grass seeds ~1 mm in size, were observed on the tops of the Hope Creek canisters

  16. The influence of the reactor pressure on the hydrodynamics in a cocurrent gas-liquid trickle-bed reactor

    NARCIS (Netherlands)

    Wammes, W.J.A.; Westerterp, K.R.

    1990-01-01

    The influence of the reactor pressure on the liquid hold-up in the trickle-flow regime and on the transition between trickle-flow and pulse-flow has been investigated in a trickle-flow column operating up to 6.0 MPa with water, and nitrogen or helium as the gas phase. The effect of the gas velocity

  17. Warm Season Storms, Floods, and Tributary Sand Inputs below Glen Canyon Dam: Investigating Salience to Adaptive Management in the Context of a 10-Year Long Controlled Flooding Experiment in Grand Canyon National Park, AZ, USA

    Science.gov (United States)

    Jain, S.; Melis, T. S.; Topping, D. J.; Pulwarty, R. S.; Eischeid, J.

    2013-12-01

    The planning and decision processes in the Glen Canyon Dam Adaptive Management Program (GCDAMP) strive to balance numerous, often competing, objectives, such as, water supply, hydropower generation, low flow maintenance, maximizing conservation of downstream tributary sand supply, endangered native fish, and other sociocultural resources of Glen Canyon National Recreation Area and Grand Canyon National Park. In this context, use of monitored and predictive information on the warm season floods (at point-to-regional scales) has been identified as lead-information for a new 10-year long controlled flooding experiment (termed the High-Flow Experiment Protocol) intended to determine management options for rebuilding and maintaining sandbars in Grand Canyon; an adaptive strategy that can potentially facilitate improved planning and dam operations. In this work, we focus on a key concern identified by the GCDAMP, related to the timing and volume of tributary sand input from the Paria and Little Colorado Rivers (located 26 and 124 km below the dam, respectively) into the Colorado River in Grand Canyon National Park. Episodic and intraseasonal variations (with links to equatorial and sub-tropical Pacific sea surface temperature variability) in the southwest hydroclimatology are investigated to understand the magnitude, timing and spatial scales of warm season floods from this relatively small, but prolific sand producing drainage of the semi-arid Colorado Plateau. The coupled variations of the flood-driven sediment input (magnitude and timing) from these two drainages into the Colorado River are also investigated. The physical processes, including diagnosis of storms and moisture sources, are mapped alongside the planning and decision processes for the ongoing experimental flood releases from the Glen Canyon Dam which are aimed at achieving restoration and maintenance of sandbars and instream ecology. The GCDAMP represents one of the most visible and widely recognized

  18. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  19. High temperature gas-cooled reactor: gas turbine application study

    International Nuclear Information System (INIS)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project

  20. Basic design and economical evaluation of Gas Turbine High Temperature Reactor 300 (GTHTR300)

    International Nuclear Information System (INIS)

    Kazuhiko, Kunitomi; Shusaku, Shiozawa; Xing, Yan

    2007-01-01

    High Temperature Gas-cooled Reactor (HTGR) combined with a direct cycle gas turbine offers one of the most promising nuclear electricity generation options after 2010. Japan Atomic Energy Agency has been engaging in the basic design and development of Gas Turbine High Temperature Reactor 300 (GTHTR300) since 2003. Costs of capital, fuel, and operation and maintenance have been estimated. The capital cost of the GTHTR300 is lower than that of the existing light water reactor (LWR) because the generation efficiency is considerably higher whereas the construction cost is lower owing to the design simplicity of the gas turbine power conversion unit and the reactor safety system. The fuel cost is shown to equal that of LWR. The operation and maintenance cost has a slight advantage due to the use of chemically inert helium coolant. In sum, the cost of electricity for the GTHTR300 is estimated to be below US 3.3 cents/kWh (4 yen/kWh), which is about two-third of that of current LWRs in Japan. The results confirm that the net power generation cost of the GTHTR300 is much lower than that of the LWR, indicating that the GTHTR300 plant consisting of small-scale reactor units can be economically competitive to the latest large-scale LWR. (authors)

  1. Gas core reactor power plants designed for low proliferation potential

    International Nuclear Information System (INIS)

    Lowry, L.L.

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF 6 and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on 233 U born from thorium. Fission product removal was continuous. Newly born 233 U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of 233 U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors

  2. 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)

  3. A gas-phase reactor powered by solar energy and ethanol for H2 production

    International Nuclear Information System (INIS)

    Ampelli, Claudio; Genovese, Chiara; Passalacqua, Rosalba; Perathoner, Siglinda; Centi, Gabriele

    2014-01-01

    In the view of H 2 as the future energy vector, we presented here the development of a homemade photo-reactor working in gas phase and easily interfacing with fuel cell devices, for H 2 production by ethanol dehydrogenation. The process generates acetaldehyde as the main co-product, which is more economically advantageous with respect to the low valuable CO 2 produced in the alternative pathway of ethanol photoreforming. The materials adopted as photocatalysts are based on TiO 2 substrates but properly modified with noble (Au) and not-noble (Cu) metals to enhance light harvesting in the visible region. The samples were characterized by BET surface area analysis, Transmission Electron Microscopy (TEM) and UV–visible Diffusive Reflectance Spectroscopy, and finally tested in our homemade photo-reactor by simulated solar irradiation. We discussed about the benefits of operating in gas phase with respect to a conventional slurry photo-reactor (minimization of scattering phenomena, no metal leaching, easy product recovery, etc.). Results showed that high H 2 productivity can be obtained in gas phase conditions, also irradiating titania photocatalysts doped with not-noble metals. - Highlights: • A gas-phase photoreactor for H 2 production by ethanol dehydrogenation was developed. • The photocatalytic behaviours of Au and Cu metal-doped TiO 2 thin layers are compared. • Benefits of operating in gas phase with respect to a slurry reactor are presented. • Gas phase conditions and use of not-noble metals are the best economic solution

  4. 78 FR 56944 - Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation

    Science.gov (United States)

    2013-09-16

    ... process waste at the Humboldt Bay ISFSI will not significantly affect the quality of the human environment... NUCLEAR REGULATORY COMMISSION [Docket No. 72-27; NRC-2011-0115] Pacific Gas and Electric Company; Humboldt Bay Independent Spent Fuel Storage Installation AGENCY: Nuclear Regulatory Commission. ACTION...

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

  11. Frequency and distribution of leakages in steam generators of gas-cooled reactors

    International Nuclear Information System (INIS)

    Bongratz, R.; Breitbach, G.; Wolters, J.

    1988-01-01

    In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm 2 , then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

  12. Gas reactor international cooperative program interim report: German Pebble Bed Reactor design and technology review

    International Nuclear Information System (INIS)

    1978-09-01

    This report describes and evaluates several gas-cooled reactor plant concepts under development within the Federal Republic of Germany (FRG). The concepts, based upon the use of a proven Pebble Bed Reactor (PBR) fuel element design, include nuclear heat generation for chemical processes and electrical power generation. Processes under consideration for the nuclear process heat plant (PNP) include hydrogasification of coal, steam gasification of coal, combined process, and long-distance chemical heat transportation. The electric plant emphasized in the report is the steam turbine cycle (HTR-K), although the gas turbine cycle (HHT) is also discussed. The study is a detailed description and evaluation of the nuclear portion of the various plants. The general conclusions are that the PBR technology is sound and that the HTR-K and PNP plant concepts appear to be achievable through appropriate continuing development programs, most of which are either under way or planned

  13. Relap5 simulation for severe accident analysis of RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Andi Sofrany Ekariansyah; Endiah P-Hastuti; Sudarmono

    2018-01-01

    The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational characteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element. (author)

  14. Risk-informed design of a pebble bed gas reactor

    International Nuclear Information System (INIS)

    Ritterbusch, Stanley; Dimitrijevic, Vesna; Simic Zdenko; Savkina Marina

    2003-01-01

    One of the major challenges to the successful deployment of new nuclear plants in the United States is the regulatory process, which is largely based on water-reactor design technology and operating experience. While ongoing and expected efforts to license new LWR designs are based primarily on current regulations, guidance, and past experience, the pre-application review of the gas-cooled Pebble Bed Modular Reactor (PBMR) has shown that efforts are being made to provide additional 'risk-informed' improvements to the licensing process. These improvements are aimed at resolving new design and regulatory issues using a plant-wide integrated evaluation method - state-of-the-art Probabilistic Risk Assessment - which addresses all significant design features and operating modes. The integrated PRA evaluation is supported by the usual deterministic design analyses, engineering judgments, and margins added to address uncertainties (i.e., defense-in-depth). The work performed for this paper was completed as part of the United States Department of Energy's Nuclear Energy Research Initiative. The purpose of this particular project was to develop the methods for a new 'highly risk-informed' design and regulatory process. In this work. PRA techniques were applied in order to provide an integrated and systematic analysis of the plant design, to quantify uncertainties and explicitly account for defense-in-depth features. This work concentrates on the application of the risk-informed principles to a new plant design such as the PBMR. The implementation example completed for this project included specification of the design configuration, use of the PRA to evaluate the design, and iterations to identify design changes that improve the overall level of safety and system reliability. This paper summarizes the new 'highly risk-informed' design process, the design of the PBMR, and the results obtained. These results, consistent with the known inherent safety features of a pebble

  15. Research and development for high temperature gas cooled reactor in Japan

    International Nuclear Information System (INIS)

    Taketani, K.

    1978-01-01

    The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

  16. High-temperature gas-cooled reactor safety-reliability program plan

    Energy Technology Data Exchange (ETDEWEB)

    1981-03-01

    The purpose of this document is to present a safety plan as part of an overall program plan for the design and development of the High Temperature Gas-Cooled Reactor (HTGR). This plan is intended to establish a logical framework for identifying the technology necessary to demonstrate that the requisite degree of public risk safety can be achieved economically. This plan provides a coherent system safety approach together with goals and success criterion as part of a unifying strategy for licensing a lead reactor plant in the near term. It is intended to provide guidance to program participants involved in producing a technology base for the HTGR that is fully responsive to safety consideration in the design, evaluation, licensing, public acceptance, and economic optimization of reactor systems.

  17. Pacific Northwest Laboratory monthly activities report, April 1965

    Energy Technology Data Exchange (ETDEWEB)

    1965-05-14

    This report discusses research at the Pacific Northwest Laboratory on topics relating to hanford production reactors. The topic deal with: reactor and material technology; reactor physics and instruments; chemistry; biology and medicine; applied mathematics; radiation protection; and test reactor and engineering services.

  18. Basic study on high temperature gas cooled reactor technology for hydrogen production

    International Nuclear Information System (INIS)

    Chang, Jong Hwa; Lee, W. J.; Lee, H. M.

    2003-01-01

    The annual production of hydrogen in the world is about 500 billion m 3 . Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

  19. LOFA and RIA analysis of the Indonesian Multipurpose research reactor RSG-GAS 1)

    International Nuclear Information System (INIS)

    Endiah Puji Hastuti; Hudi Hastowo; Iman Kuntoro

    1999-01-01

    Investigation on accident of the Indonesian Multipurpose research reactor RSG-GAS has been performed by computer simulation technique. Two groups of transients were considered, namely transient due to loss of primary cooling system (LOFA) and power excursion due to reactivity insertion (RIA). In such a transient condition, the Common Mode Failure (CMF) is considered and it will induce a situation so called unprotected transient or Anticipated Transient Without Scram (ATWS). RELAP5, PARET-ANL and EUREKA-2RR computer packages have been applied for these analyses. Simulations result done using these computer packages showed that in the occurrence of LOFA and RIA, failure on fuel elements is limited to the region with the highest power factor. (author)

  20. Cavity temperature and flow characteristics in a gas-core test reactor

    Science.gov (United States)

    Putre, H. A.

    1973-01-01

    A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

  1. Hydrogeology of Middle Canyon, Oquirrh Mountains, Tooele County, Utah

    Science.gov (United States)

    Gates, Joseph Spencer

    1963-01-01

    Geology and climate are the principal influences affecting the hydrology of Middle Canyon, Tooele County, Utah. Reconnaissance in the canyon indicated that the geologic influences on the hydrology may be localized; water may be leaking through fault and fracture zones or joints in sandstone and through solution openings in limestone of the Oquirrh formation of Pennsylvanian and Permian age. Surficial deposits of Quaternary age serve as the main storage material for ground water in the canyon and transmit water from the upper canyon to springs and drains at the canyon mouth. The upper canyon is a more important storage area than the lower canyon because the surficial deposits are thicker, and any zones of leakage in the underlying bedrock of the upper canyon probably would result in greater leakage than would similar outlets in the lower canyon.The total annual discharge from Middle Canyon, per unit of precipitation, decreased between 1910 and 1939. Similar decreases occurred in Parleys Canyon in the nearby Wasatch Range and in other drainage basins in Utah, and it is likely that most of the decrease in discharge from Middle Canyon and other canyons in Utah is due to a change in climate.Chemical analyses of water showed that the high content of sulfate and other constituents in the water from the Utah Metals tunnel, which drains into Middle Canyon, does not have a significant effect on water quality at the canyon mouth. This suggests that much of the tunnel water is lost from the channel by leakage, probably in the upper canyon, during the dry part of the year.Comparison of the 150 acre-feet of water per square mile of drainage area discharged by Middle Canyon in 1947 with the 623 and 543 acre-feet per square mile discharged in 1948 by City Creek and Mill Creek Canyons, two comparable drainage basins in the nearby Wasatch Range, also suggests that there is leakage in Middle Canyon.A hydrologic budget of the drainage basin results in an estimate that about 3,000 acre

  2. Fuel Summary for Peach Bottom Unit 1 High-Temperature Gas-Cooled Reactor Cores 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Karel I. Kingrey

    2003-04-01

    This fuel summary report contains background and summary information for the Peach Bottom Unit 1, High-Temperature, Gas-Cooled Reactor Cores 1 and 2. This report contains detailed information about the fuel in the two cores, the Peach Bottom Unit 1 operating history, nuclear parameters, physical and chemical characteristics, and shipping and storage canister related data. The data in this document have been compiled from a large number of sources and are not qualified beyond the qualification of the source documents. This report is intended to provide an overview of the existing data pertaining to spent fuel management and point to pertinent reference source documents. For design applications, the original source documentation must be used. While all referenced sources are available as records or controlled documents at the Idaho National Engineering and Environmental Laboratory (INEEL), some of the sources were marked as informal or draft reports. This is noted where applicable. In some instances, source documents are not consistent. Where they are known, this document identifies those instances and provides clarification where possible. However, as stated above, this document has not been independently qualified and such clarifications are only included for information purposes. Some of the information in this summary is available in multiple source documents. An effort has been made to clearly identify at least one record document as the source for the information included in this report.

  3. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1986-09-01

    Two sodium cooled reactors are currently being operated in the United States of America for the US Department of Energy. These are Experimental Breeder Reactor 11, EBR-11, and the Fast Flux Test Facility, FFTF. EBR-11 is located near Idaho Falls, Idaho, and the FFTF is near Richland, Washington. These reactors are currently engaged in a wide range of testing including fuels and materials tests, and plant system performance and safety development. The US DOE program also includes designs of a next generation sodium cooled power reactor. The FFTF and EBR-11 communities are providing input to these designs. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  4. Methods and devices prepared to eliminate activation and fission products from PEC reactor cover gas

    International Nuclear Information System (INIS)

    Caponetti, R.; Gherardi, G.; Petrazzuolo, F.

    1987-01-01

    The major effort made in Italy for the development of fast nuclear reactor is concentrated in the PEC reactor, whose construction is now in the completion stage. The PEC reactor (Prova Elementi di Combustibile - Fuel Element Testing ) is a sodium-cooled reactor with a power rating of 120 MWt, being built for the purpose of studying the behavior of fuel elements under thermal and neutronic conditions similar to those of fast reactor power stations, whit particular attention to safety aspects. The PEC reactor represents a research instrument particularly suitable for studies and experiments in the following fields: performances of the fuel element and its economical optimization (also with the possibility of testing fuel elements not necessarily based on mixed oxides); experiments in the safety field, not only referred to fuel elements, but also to plant subsystems. The experimental program will cover the research of the limit conditions of the typical parameters, such as cladding temperature, linear power, radiation rate, etc. PEC will also allow researches on new-concept fuel elements and thermal, hydraulic and power transients and cycles foreseen in the commercial power plants under normal, upset and emergency conditions. A number of the solutions regarding the PEC reactor and preparatory approaches to its operation are reported in this paper. In particular the following items are discussed: a description of three cover-gas circuits present in the reactor; an estimate of the contamination conditions foreseen under operating conditions; a description of the equipment for the purification of the cover gas and relative operating conditions. There are three cover-gas circuits present in the PEC reactor. They concern the following sodium circuits: primary reactor, primary emergency reactor and sodium purification primary reactor; secondary reactor, test channel and emergency reactor; primary test channel

  5. Catalytic and Noncatalytic Conversion of Methane to Olefins and Synthesis Gas in an AC Parallel Plate Discharge Reactor

    Directory of Open Access Journals (Sweden)

    Mohammad Ali Khodagholi

    2013-01-01

    Full Text Available Direct conversion of methane to ethylene, acetylene, and synthesis gas at ambient pressure and temperature in a parallel plate discharge reactor was investigated. The experiments were carried out using a quartz reactor of outer diameter of 9 millimeter and a driving force of ac current of 50 Hz. The input power to the reactor to establish a stable gas discharge varied from 9.6 to maximum 15.3 watts (w. The effects of ZSM5, Fe–ZSM5, and Ni–ZSM5 catalysts combined with corona discharge for conversion of methane to more valued products have been addressed. It was found that in presence or absence of a catalyst in gas discharge reactor, the rate of methane and oxygen conversion increased upon higher input power supplied to the reactor. The effect of Fe–ZSM5 catalyst combined with gas discharge plasma yields C2 hydrocarbons up to 21.9%, which is the highest productions of C2 hydrocarbons in this work. The effect of combined Ni–ZSM5 and gas discharge plasma was mainly production of synthesis gas. The advantage of introducing ZSM5 to the plasma zone was increase in synthesis gas and acetylene production. The highest energy efficiency was 0.22 mmol/kJ, which belongs to lower rate of energy injection to the reactor.

  6. Use of thorium for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

    2017-07-01

    The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)

  7. Microscopical examination of carbon deposits formed in the Windscale advanced gas cooled reactor

    International Nuclear Information System (INIS)

    Livesey, D.J.; Chatwin, W.H.; Pearce, J.H.

    1980-12-01

    Methods are described of sampling and examining carbon deposits on fuel cladding in the Windscale advanced gas-cooled reactor. Deposition is observed on fuel cladding in both the reactor core and experimental loops in carbon dioxide coolants containing various amounts of carbon monoxide and methane. Deposit distribution over the cladding surface indicated that nucleation is dependent on local surface conditions. Microscopical examination showed that deposit thickness increases by carbon filament growth into the coolant gas stream and that the process can be markedly influenced by metallic impurities. There is evidence that nickel can play a particularly significant role in deposition in loop experiments but similar effects have not been observed in the reactor core. (author)

  8. Gas/liquid separator for BWR type reactor

    International Nuclear Information System (INIS)

    Soma, Naoshi; Akimoto, Seiichi; Yokoyama, Iwao.

    1993-01-01

    A two phase gas/liquid flow generated at a heating portion of a nuclear reactor is swirled by inlet vanes. The phase gas/liquid flow uprises as a vortex flow in a vortex cylinder, and a liquid phase of a high density gathers at the outer circumference of the vortex cylinder. The liquid phase gathered at the outer circumference is collected at the inlet of a discharge flow channel which protrude into the vortex cylinder and in a three-step structure, and introduced into a recycling liquid phase passing through the discharge flow channel for liquid phase. There is provided a structure that separated liquid collected at the lowermost state in the inlet of the three-step discharge flow channel inlet descends in the discharge flow channel, then uprises in an uprising flow channel and is introduced into the recycling liquid phase by way of a discharge flow channel exit. The height of the discharge flow channel exit is determined equal to that of a liquid level of the recycling liquid phase during rated operation of the reactor. Accordingly, even in a case where the liquid level in the recycling liquid phase is lowered, the liquid level of the uprising flow channel is kept equal to that during rated operation. (I.N.)

  9. Presentation summary: Gas Turbine - Modular Helium Reactor (GT-MHR)

    International Nuclear Information System (INIS)

    2001-01-01

    Numerous prototypes and demonstration plants have been constructed and operated beginning with the Dragon plant in the early 1960s. The MHTGR was the U.S. developed modular plant and underwent pre application review by NRC. The GT-MHR represents a further refinement on this concept with the steam cycle being replaced by a closed loop gas turbine (Brayton) cycle. Modular gas reactors and the GT-MHR represent a fundamental shift in reactor design and safety philosophy. The reactor system is contained in a 3 vessel, side-by-side arrangement. The reactor and a shutdown cooling system are in one vessel, and the gas turbine based power conversion system, including the generator, in a second parallel vessel. A more detailed look at the system shows the compact arrangement of gas turbine, compressors, recuperator, heat exchanges, and generator. Fueled blocks are stacked in three concentric rings with inert graphite blocks making up the inner and outer reflectors. Operating control rods are located outside the active core while startup control rods and channels for reserve shutdown pellets are located near the core center. Ceramic coated fuel is the key to the GT-MHR's safety and economics. A kernel of Uranium oxycarbide (or UO 2 ) is placed in a porous carbon buffer and then encapsulated in multiple layers of pyrolytic carbon and silicon carbide. These micro pressure vessels withstand internal pressures of up to 2,000 psi and temperatures of nearly 2,000 C providing extremely resilient containment of fission products under both normal operating and accident conditions. The fuel particles are blended in carbon pitch, forming fuel rods, and then loaded into holes within large graphite fuel elements. Fuel elements are stacked to form the core. Fuel particle testing in has repeatedly demonstrated the high temperature resilience of coated particle fuel to temperature approaching 2,000 C. As an conservative design goal, GT-MHR has been sized to keep maximum fuel temperatures

  10. Advanced In-Core Fuel Cycles for the Gas Turbine-Modular Helium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto

    2006-04-15

    Amid generation IV of nuclear power plants, the Gas Turbine - Modular Helium Reactor, designed by General Atomics, is the only core with an energy conversion efficiency of 50%; the safety aspects, coupled to construction and operation costs lower than ordinary Light Water Reactors, renders the Gas Turbine - Modular Helium reactor rather unequaled. In the present studies we investigated the possibility to operate the GT-MHR with two types of fuels: LWRs waste and thorium; since thorium is made of only fertile {sup 232}Th, we tried to mix it with pure {sup 233}U, {sup 235}U or {sup 239}Pu; ex post facto, only uranium isotopes allow the reactor operation, that induced us to examine the possibility to use a mixture of uranium, enriched 20% in {sup 235}U, and thorium. We performed all calculations by the MCNP and MCB codes, which allowed to model the reactor in a very detailed three-dimensional geometry and to describe the nuclides transmutation in a continuous energy approach; finally, we completed our studies by verifying the influence of the major nuclear data libraries, JEFF, JENDL and ENDF/B, on the obtained results.

  11. Feasibility study of full-reactor gas core demonstration test

    Science.gov (United States)

    Kunze, J. F.; Lofthouse, J. H.; Shaffer, C. J.; Macbeth, P. J.

    1973-01-01

    Separate studies of nuclear criticality, flow patterns, and thermodynamics for the gas core reactor concept have all given positive indications of its feasibility. However, before serious design for a full scale gas core application can be made, feasibility must be shown for operation with full interaction of the nuclear, thermal, and hydraulic effects. A minimum sized, and hence minimum expense, test arrangement is considered for a full gas core configuration. It is shown that the hydrogen coolant scattering effects dominate the nuclear considerations at elevated temperatures. A cavity diameter of somewhat larger than 4 ft (122 cm) will be needed if temperatures high enough to vaporize uranium are to be achieved.

  12. Gas-Liquid Two-Phase Flows Through Packed Bed Reactors in Microgravity

    Science.gov (United States)

    Motil, Brian J.; Balakotaiah, Vemuri

    2001-01-01

    The simultaneous flow of gas and liquid through a fixed bed of particles occurs in many unit operations of interest to the designers of space-based as well as terrestrial equipment. Examples include separation columns, gas-liquid reactors, humidification, drying, extraction, and leaching. These operations are critical to a wide variety of industries such as petroleum, pharmaceutical, mining, biological, and chemical. NASA recognizes that similar operations will need to be performed in space and on planetary bodies such as Mars if we are to achieve our goals of human exploration and the development of space. The goal of this research is to understand how to apply our current understanding of two-phase fluid flow through fixed-bed reactors to zero- or partial-gravity environments. Previous experiments by NASA have shown that reactors designed to work on Earth do not necessarily function in a similar manner in space. Two experiments, the Water Processor Assembly and the Volatile Removal Assembly have encountered difficulties in predicting and controlling the distribution of the phases (a crucial element in the operation of this type of reactor) as well as the overall pressure drop.

  13. Strategic guidelines for street canyon geometry to achieve sustainable street air quality

    Energy Technology Data Exchange (ETDEWEB)

    Chan, Andy T.; So, Ellen S.P.; Samad, Subash C. [Hong Kong Univ., Dept. of Mechanical Engineering, Hong Kong (China)

    2001-08-01

    This paper is concerned with the motion of air within the urban street canyon and is directed towards a deeper understanding of pollutant dispersion with respect to various simple canyon geometries and source positions. Taking into account the present days typical urban configurations, three principal flow regimes 'isolated roughness flow', 'skimming flow' and 'wake interference flow' (Boundary Layer Climates, 2nd edition, Methuen, London) and their corresponding pollutant dispersion characteristics are studied for various canopies aspect ratios, namely relative height (h{sub 2}/H{sub 1}), canyon height to width ratio (h/w) and canyon length to height ratio (l/h). A field-size canyon has been analysed through numerical simulations using the standard k-{sup {epsilon}} turbulence closure model. It is found that the pollutant transport and diffusion is strongly dependent upon the type of flow regime inside the canyon and exchange between canyon and the above roof air. Some rules of thumbs have been established to get urban canyon geometries for efficient dispersion of pollutants. (Author)

  14. Corals, Canyons, and Conservation: Science Based Fisheries Management Decisions in the Eastern Bering Sea

    Directory of Open Access Journals (Sweden)

    Steve A. MacLean

    2017-05-01

    Full Text Available When making science matter for conservation, marine conservation practitioners, and managers must be prepared to make the appropriate decision based on the results of the best available science used to inform it. For nearly a decade, many stakeholders encouraged the North Pacific Fishery Management Council to enact protections for deep-sea corals in several canyons in the Eastern Bering Sea slope. In 2014, at the request of the Council, the National Marine Fisheries Service, Alaska Fisheries Science Center conducted a strip-transect survey along the Eastern Bering Sea slope to validate the results of a model predicting the occurrence of deep-sea coral habitat. More than 250,000 photos were analyzed to estimate coral, sponge, and sea whip abundance, distribution, height, and vulnerability to anthropogenic damage. The results of the survey confirmed that coral habitat and occurrence was concentrated around Pribilof Canyon and the adjacent slope. The results also confirmed that the densities of corals in the Eastern Bering Sea were low, even where they occurred. After reviewing the best available scientific information, the Council concluded that there is no scientific evidence to suggest that deep-sea corals in the Eastern Bering Sea slope or canyons are at risk from commercial fisheries under the current management structure, and that special protections for deep-sea corals were not warranted.

  15. Preliminary Sensitivity Study on Gas-Cooled Reactor for NHDD System Using MARS-GCR

    International Nuclear Information System (INIS)

    Lee, Seung Wook; Jeong, Jae Jun; Lee, Won Jae

    2005-01-01

    A Gas-Cooled Reactor (GCR) is considered as one of the most outstanding tools for a massive hydrogen production without CO 2 emission. Till now, two types of GCR are regarded as a viable nuclear reactor for a hydrogen production: Prismatic Modular Reactor (PMR), Pebble Bed Reactor (PBR). In this paper, a preliminary sensitivity study on two types of GCR is carried out by using MARS-GCR to find out the effect on the peak fuel and reactor pressure vessel (RPV) temperature, with varying the condition of a reactor inlet, outlet temperature, and system pressure for both PMR and PBR

  16. PREDICTION OF GAS HOLD-UP IN A COMBINED LOOP AIR LIFT FLUIDIZED BED REACTOR USING NEWTONIAN AND NON-NEWTONIAN LIQUIDS

    Directory of Open Access Journals (Sweden)

    Sivakumar Venkatachalam

    2011-09-01

    Full Text Available Many experiments have been conducted to study the hydrodynamic characteristics of column reactors and loop reactors. In this present work, a novel combined loop airlift fluidized bed reactor was developed to study the effect of superficial gas and liquid velocities, particle diameter, fluid properties on gas holdup by using Newtonian and non-Newtonian liquids. Compressed air was used as gas phase. Water, 5% n-butanol, various concentrations of glycerol (60 and 80% were used as Newtonian liquids, and different concentrations of carboxy methyl cellulose aqueous solutions (0.25, 0.6 and 1.0% were used as non-Newtonian liquids. Different sizes of spheres, Bearl saddles and Raschig rings were used as solid phases. From the experimental results, it was found that the increase in superficial gas velocity increases the gas holdup, but it decreases with increase in superficial liquid velocity and viscosity of liquids. Based on the experimental results a correlation was developed to predict the gas hold-up for Newtonian and non-Newtonian liquids for a wide range of operating conditions at a homogeneous flow regime where the superficial gas velocity is approximately less than 5 cm/s

  17. 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)

  18. 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)

  19. Study on the incore fuel management of RSG-GAS reactor

    International Nuclear Information System (INIS)

    Kuntjoro, S.; Sutiarso, K.; Praptoriadi, G.

    1995-01-01

    The RSG-GAS reactor was designed to work at nominal thermal power of 30 MW with average thermal neutron flux of 2.10 -14 cm -2 s -1 . Forty fuel assemblies and eight control assemblies are expected to form the typical working core (TWC) so that the pattern of the in-core fel management could be carried out properly. One of the requested safety conditions is that the reactor could still be shut down even if the control assemblies with highest negative reactivity were in failure work properly. It is not possible, with respect to the prerequested safety condition, to construct directly a typical working core. Utilizing the IAFUEL code program, a series of transition cores (the cores before reaching the TWC) were set up. The TWC is expected to be reached after eight transition cores. (author)

  20. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  1. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core

  2. Proliferation resistance assessment of high temperature gas reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chikamatsu N, M. A. [Instituto Tecnologico y de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: midori.chika@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  3. Proliferation resistance assessment of high temperature gas reactors

    International Nuclear Information System (INIS)

    Chikamatsu N, M. A.; Puente E, F.

    2014-10-01

    The Generation IV International Forum has established different objectives for the new generation of reactors to accomplish. These objectives are focused on sustain ability, safety, economics and proliferation resistance. This paper is focused on how the proliferation resistance of the High Temperature Gas Reactors (HTGR) is assessed and the advantages that these reactors present currently. In this paper, the focus will be on explaining why such reactors, HTGR, can achieve the goals established by the GIF and can present a viable option in terms of proliferation resistance, which is an issue of great importance in the field of nuclear energy generation. The reason why the HTGR are being targeted in this writing is that these reactors are versatile, and present different options from modular reactors to reactors with the same size as the ones that are being operated today. Besides their versatility, the HTGR has designed features that might improve on the overall sustain ability of the nuclear reactors. This is because the type of safety features and materials that are used open up options for industrial processes to be carried out; cogeneration for instance. There is a small section that mentions how HTGR s are being developed in the international sector in order to present the current world view in this type of technology and the further developments that are being sought. For the proliferation resistance section, the focus is on both the intrinsic and the extrinsic features of the nuclear systems. The paper presents a comparison between the features of Light Water Reactors (LWR) and the HTGR in order to be able to properly compare the most used technology today and one that is gaining international interest. (Author)

  4. Combined Mobile In Situ and Remote Sensing Investigation of the Aliso Canyon Natural Gas Leak in the Los Angeles Basin, California

    Science.gov (United States)

    Hall, J. L.; Leifer, I.; Tratt, D. M.; Melton, C.; Frash, J.

    2016-12-01

    The Aliso Canyon natural gas leak in the San Fernando Valley in Los Angeles, California was a major disruptive event whose societal impacts have continued well after the event itself ended, yet fortunately did not involve highly toxic gases. Chemical releases can have serious consequences for ecosystems, societies, and human health. Mitigating their destructive impacts relies on identification and mapping, monitoring, and trajectory forecasting. Improvements in the accuracy of such transport modeling capabilities can significantly improve the effectiveness of disaster response activities. Simultaneous plume characterization data were collected by the Mobile Infrared Sensor for Tactical Incident Response (MISTIR) and AutoMObile trace Gas (AMOG) Surveyor, two instrumented vehicles traveling in convoy. Surface vehicles have advantages over airplanes in terms of simpler logistics, such as not being limited by controlled airspace which is a major issue in Los Angeles, and ability to deploy rapidly. Moreover, it is the surface concentration that impacts human health and determines ecological damage. Fusion of the resulting correlative surface in situ observations and thermal-infrared spectroscopic column observations allowed both lateral and temporal plume characterization to derive emissions and to characterize the confining effect of topography on plume dispersion. Although a straightforward Gaussian plume inversion approach based on surface data yields an emission estimate with reasonable fidelity, it required assumptions of vertical profile and topographic influence that were validated by the column spectroscopic observations. Topographic factors within the Los Angeles Basin, including the Aliso Canyon locale, strongly influence transport processes. This situation challenges the predictive skill of numerical transport models that are used to assist the evacuation of at-risk communities, for example in the case of a refinery fire. This study demonstrated the utility

  5. Power Conversion Study for High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Chang Oh; Richard Moore; Robert Barner

    2005-01-01

    The Idaho National Laboratory (INL) is investigating a Brayton cycle efficiency improvement on a high temperature gas-cooled reactor (HTGR) as part of Generation-IV nuclear engineering research initiative. There are some technical issues to be resolved before the selection of the final design of the high temperature gas cooled reactor, called as a Next Generation Nuclear Plant (NGNP), which is supposed to be built at the INEEL by year 2017. The technical issues are the selection of the working fluid, direct vs. indirect cycle, power cycle type, the optimized design in terms of a number of intercoolers, and others. In this paper, we investigated a number of working fluids for the power conversion loop, direct versus indirect cycle, the effect of intercoolers, and other thermal hydraulics issues. However, in this paper, we present part of the results we have obtained. HYSYS computer code was used along with a computer model developed using Visual Basic computer language

  6. Integrated Outcrop and Subsurface Studies of the Interwell Environment of Carbonate Reservoirs: Clear Fork (Leonaradian Age) Reservoirs, West Texas and New Mexico

    International Nuclear Information System (INIS)

    Lucia, F.J.; Ruppel, S.C.

    2001-01-01

    Characterization of cycle and facies architecture on lower Clear Fork and lowermost upper Clear Fork equivalent outcrops in Apache Canyon of Sierra Diablo was complete. The focus of detailed study in Apache Canyon has been the upper Clear Fork section because this interval contains the productive interval in South Wasson field, the preliminary subsurface study area. Parts of three high-frequency sequences (HFS), each 60 to 100 ft thick, are present on the south wall of Apache Canyon. HFS's display an upper-deepening or backstepping pattern associated with longer-term sea level rise. Each HFS is composed of upward-shallowing cycles whose thickness, facies composition, and continuity vary within and between HFS's

  7. Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Araj, K.

    1983-01-01

    The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

  8. On the impact of trees on dispersion processes of traffic emissions in street canyons

    NARCIS (Netherlands)

    Gromke, C.B.; Ruck, B.

    2009-01-01

    Wind-tunnel studies of dispersion processes of traffic exhaust in urban street canyons with tree planting were performed and tracer gas concentrations using electron capture detection (ECD) and flow fields using laser Doppler velocimetry (LDV) were measured. It was found that tree planting reduces

  9. FUEL BURN-UP CALCULATION FOR WORKING CORE OF THE RSG-GAS RESEARCH REACTOR AT BATAN SERPONG

    Directory of Open Access Journals (Sweden)

    Tukiran Surbakti

    2017-12-01

    Full Text Available The neutronic parameters are required in the safety analysis of the RSG-GAS research reactor. The RSG-GAS research reactor, MTR (Material Testing Reactor type is used for research and also in radioisotope production. RSG-GAS has been operating for 30 years without experiencing significant obstacles. It is managed under strict requirements, especially fuel management and fuel burn-up calculations. The reactor is operated under the supervision of the Regulatory Body (BAPETEN and the IAEA (International Atomic Energy Agency. In this paper, the experience of managing RSG-GAS core fuels will be discussed, there are hundred possibilities of fuel placements on the reactor core and the strategy used to operate the reactor will be crucial. However, based on strict calculation and supervision, there is no incorrect placement of the fuels in the core. The calculations were performed on working core by using the WIMSD-5B computer code with ENDFVII.0 data file to generate the macroscopic cross-section of fuel and BATAN-FUEL code were used to obtain the neutronic parameter value such as fuel burn-up fractions. The calculation of the neutronic core parameters of the RSG-GAS research reactor was carried out for U3Si2-Al fuel, 250 grams of mass, with an equilibrium core strategy. The calculations show that on the last three operating cores (T90, T91, T92, all fuels meet the safety criteria and the fuel burn-up does not exceed the maximum discharge burn-up of 59%. Maximum fuel burn-up always exists in the fuel which is close to the position of control rod.

  10. Pyrolysis of rice husk and corn stalk in auger reactor:Part 1. Characterization of char and gas at various temperatures

    OpenAIRE

    Yu, Yang; Yang, Yang; Cheng, Zhicai; Blanco, Paula H.; Liu, Ronghou; Bridgwater, A.V.; Cai, Junmeng

    2016-01-01

    In this study, rice husk and corn stalk have been pyrolyzed in an auger pyrolysis reactor at pyrolysis temperatures of 350, 400, 450, 500, 550, and 600 °C in order to investigate the effect of the pyrolysis temperature on the pyrolysis performance of the reactor and physicochemical properties of pyrolysis products (this paper focuses on char and gas). The results have shown that the pyrolysis temperature significantly affects the mass yields and properties of the pyrolysis products. The mass ...

  11. A numerical analysis of pollutant dispersion in street canyon: influence of the turbulent Schmidt number

    Directory of Open Access Journals (Sweden)

    Bouabdellah Abed

    2017-12-01

    Full Text Available Realizing the growing importance and availability of motor vehicles, we observe that the main source of pollution in the street canyons comes from the dispersion of automobile engine exhaust gas. It represents a substantial effect on the micro-climate conditions in urban areas. Seven idealized-2D building configurations are investigated by numerical simulations. The turbulent Schmidt number is introduced in the pollutant transport equation in order the take into account the proportion between the rate of momentum turbulent transport and the mass turbulent transport by diffusion. In the present paper, we attempt to approach the experimental test results by adjusting the values of turbulent Schmidt number to its corresponding application. It was with interest that we established this link for achieving our objectives, since the numerical results agree well with the experimental ones. The CFD code ANSYS CFX, the k, e and the RNGk-e models of turbulence have been adopted for the resolutions. From the simulation results, the turbulent Schmidt number is a range of 0.1 to 1.3 that has some effect on the prediction of pollutant dispersion in the street canyons. In the case of a flat roof canyon configuration (case: runa000, appropriate turbulent Schmidt number of 0.6 is estimated using the k-epsilon model and of 0.5 using the RNG k-e model.

  12. Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

    NARCIS (Netherlands)

    Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

    1987-01-01

    The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

  13. Optimisation of gas-cooled reactors with the aid of mathematical computers

    Energy Technology Data Exchange (ETDEWEB)

    Margen, P H

    1959-04-15

    Reactor optimisation is the task of finding the combination of values of the independent variables in a reactor design producing the lowest cost of electricity. In a gas-cooled reactor the number of independent variables is particularly large and the optimisation process is, therefore, laborious. The present note describes a procedure for performing the entire optimisation procedure with the aid of a mathematical computer in a single operation, thus saving time for the design staff. Detailed equations and numerical constants are proposed for the thermal and cost relations involved. The reactor physics equations, on the other hand are merely stated as general functions of the relevant variables. The task of expressing these functions as detailed equations will be covered by separate documents prepared by the reactor physics department.

  14. Optimisation of gas-cooled reactors with the aid of mathematical computers

    International Nuclear Information System (INIS)

    Margen, P.H.

    1959-04-01

    Reactor optimisation is the task of finding the combination of values of the independent variables in a reactor design producing the lowest cost of electricity. In a gas-cooled reactor the number of independent variables is particularly large and the optimisation process is, therefore, laborious. The present note describes a procedure for performing the entire optimisation procedure with the aid of a mathematical computer in a single operation, thus saving time for the design staff. Detailed equations and numerical constants are proposed for the thermal and cost relations involved. The reactor physics equations, on the other hand are merely stated as general functions of the relevant variables. The task of expressing these functions as detailed equations will be covered by separate documents prepared by the reactor physics department

  15. Scientific Objectives of the Gulf of Mexico Gas Hydrate JIP Leg II Drilling

    Energy Technology Data Exchange (ETDEWEB)

    Jones, E. (Chevron); Latham, T. (Chevron); McConnell, D. (AOA Geophysics); Frye, M. (Minerals Management Service); Hunt, J. (Minerals Management Service); Shedd, W. (Minerals Management Service); Shelander, D. (Schlumberger); Boswell, R.M. (NETL); Rose, K.K. (NETL); Ruppel, C. (USGS); Hutchinson, D. (USGS); Collett, T. (USGS); Dugan, B. (Rice University); Wood, W. (Naval Research Laboratory)

    2008-05-01

    The Gulf of Mexico Methane Hydrate Joint Industry Project (JIP) has been performing research on marine gas hydrates since 2001 and is sponsored by both the JIP members and the U.S. Department of Energy. In 2005, the JIP drilled the Atwater Valley and Keathley Canyon exploration blocks in the Gulf of Mexico to acquire downhole logs and recover cores in silt- and clay-dominated sediments interpreted to contain gas hydrate based on analysis of existing 3-D seismic data prior to drilling. The new 2007-2009 phase of logging and coring, which is described in this paper, will concentrate on gas hydrate-bearing sands in the Alaminos Canyon, Green Canyon, and Walker Ridge protraction areas. Locations were selected to target higher permeability, coarser-grained lithologies (e.g., sands) that have the potential for hosting high saturations of gas hydrate and to assist the U.S. Minerals Management Service with its assessment of gas hydrate resources in the Gulf of Mexico. This paper discusses the scientific objectives for drilling during the upcoming campaign and presents the results from analyzing existing seismic and well log data as part of the site selection process. Alaminos Canyon 818 has the most complete data set of the selected blocks, with both seismic data and comprehensive downhole log data consistent with the occurrence of gas hydrate-bearing sands. Preliminary analyses suggest that the Frio sandstone just above the base of the gas hydrate stability zone may have up to 80% of the available sediment pore space occupied by gas hydrate. The proposed sites in the Green Canyon and Walker Ridge areas are also interpreted to have gas hydrate-bearing sands near the base of the gas hydrate stability zone, but the choice of specific drill sites is not yet complete. The Green Canyon site coincides with a 4-way closure within a Pleistocene sand unit in an area of strong gas flux just south of the Sigsbee Escarpment. The Walker Ridge site is characterized by a sand

  16. Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

    International Nuclear Information System (INIS)

    Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

    2003-01-01

    A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

  17. Optimization of a water-gas shift reactor over a Pt/ceria/alumina monolith

    Energy Technology Data Exchange (ETDEWEB)

    Quiney, A.S.; Germani, G.; Schuurman, Y. [Institut de Recherches sur la Catalyse-CNRS, 2 Avenue A. Einstein, 69626 Villeurbanne (France)

    2006-10-06

    The water-gas shift (WGS) reaction is an important step in the purification of hydrogen for fuel cells. It lowers the carbon monoxide content and produces extra hydrogen. The constraints of automotive applications render the commercial WGS catalysts unsuitable. Pt/ceria catalysts are cited as promising catalysts for onboard applications as they are highly active and non-pyrophoric. This paper reports on a power law rate expression for a Pt/CeO{sub 2}/Al{sub 2}O{sub 3} catalyst. This rate equation is used to compare different reactor configurations for an onboard water-gas shift reactor. A one-dimensional heterogeneous model that accounts for the interfacial and intraparticle gradients has been used to optimize a dual stage adiabatic monolith reactor. (author)

  18. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  19. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  20. Sensitivity analysis of power excursion in RSG-GAS reactor due to reactivity insertion

    International Nuclear Information System (INIS)

    Pinem, Surian; Sembiring, Tagor Malem

    2002-01-01

    Reactor kinetics has a very important role in reactor operation safety and nuclear reactor control. One of the important aspects in reactor kinetics is power behavior as function of time due to chain reaction in the core. The calculation was performed using kinetic equation and feedback reactivity and evaluated using static power coefficient. Analysis was performed for oxide 250 g, silicide 250 g and silicide 300 g fuel elements with insertion of positive reactivity, negative reactivity and reactivity close to delay neutron fraction. The calculation of power excursion sensitivity showed that the insertion of 0,5 % Δk/k, in the fuel element of silicide 300 g is bigger 5 % than the one of oxide 250 g or silicide 250 g. If inserted by - 1,2 % Δk/k, there is no change among three fuel elements. Therefore, in kinetic point of view, it is showed there is no significant influence in the RSG-GAS reactor operation safety is the current core of oxide 250 g is converted to silicide 250 g or to silicide 300 g

  1. Separations canyon decontamination facilities

    International Nuclear Information System (INIS)

    Hershey, J.H.

    1975-01-01

    Highly radioactive process equipment is decontaminated at the Savannah River Plant in specially equipped areas of the separations canyon building so that direct mechanical repairs or alterations can be made. Using these facilities it is possible to decontaminate and repair equipment such as 10- x 11-ft storage tanks, 8- x 8-ft batch evaporator pots and columns, 40-in. Bird centrifuges, canyon pumps and agitators, and various canyon piping systems or ''jumpers.'' For example, centrifuge or evaporator pots can be decontaminated and rebuilt for about 60 percent of the 1974 replacement cost. The combined facilities can decontaminate and repair 6 to 10 pieces of major equipment per year. Decontamination time varies with type of equipment and radioactivity levels encountered

  2. Separations canyon decontamination facilities

    International Nuclear Information System (INIS)

    Hershey, J.H.

    1975-05-01

    Highly radioactive process equipment is decontaminated at the Savannah River Plant in specially equipped areas of the separations canyon buildings so that direct mechanical repairs or alterations can be made. Using these facilities it is possible to decontaminate and repair equipment such as 10- x 11-ft storage tanks, 8- x 8-ft batch evaporator pots and columns, 40-in. Bird centrifuges, canyon pumps and agitators, and various canyon piping systems or ''jumpers.'' For example, centrifuge or evaporator pots can be decontaminated and rebuilt for about 60 percent of the 1974 replacement cost. The combined facilities can decontaminate and repair 6 to 10 pieces of major equipment per year. Decontamination time varies with type of equipment and radioactivity levels encountered. (U.S.)

  3. 9th Pacific Basin Nuclear Conference. Nuclear energy, science and technology - Pacific partnership. Proceedings Volume 2

    International Nuclear Information System (INIS)

    1994-04-01

    The theme of the 9th Pacific Basin Nuclear Conference held in Sydney from 1-6 May 1994, embraced the use of atom in energy production and in science and technology. The focus was on selected topics of current and on-going interest to countries around the Pacific Basin. The two-volume proceedings include both invited and contributed papers which have been indexed separately. This document, Volume 2 covers the following topics: education and training in Nuclear Science, public acceptance, nuclear safety and radiation protection, nuclear fuel resources and their utilisation, research reactors, cyclotrons and accelerators. refs., tabs., figs., ills

  4. The modular high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Lutz, D.E.; Lipps, A.J.

    1984-01-01

    Due to relatively high operating temperatures, the gas-cooled reactor has the potential to serve a wide variety of energy applications. This paper discusses the energy applications which can be served by the modular HTGR, the magnitude of the potential markets, and the HTGR product cost incentives relative to fossil fuel competition. Advantages of the HTGR modular systems are presented along with a description of the design features and performance characteristics of the current reference HTGR modular systems

  5. Thermal-hydraulic code selection for modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Komen, E M.J.; Bogaard, J.P.A. van den

    1995-06-01

    In order to study the transient thermal-hydraulic system behaviour of modular high temperature gas-cooled reactors, the thermal-hydraulic computer codes RELAP5, MELCOR, THATCH, MORECA, and VSOP are considered at the Netherlands Energy Research Foundation ECN. This report presents the selection of the most appropriate codes. To cover the range of relevant accidents, a suite of three codes is recommended for analyses of HTR-M and MHTGR reactors. (orig.).

  6. 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.)

  7. 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)

  8. 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.)

  9. 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)

  10. Modular High Temperature Gas-Cooled Reactor heat source for coal conversion

    International Nuclear Information System (INIS)

    Schleicher, R.W. Jr.; Lewis, A.C.

    1992-09-01

    In the industrial nations, transportable fuels in the form of natural gas and petroleum derivatives constitute a primary energy source nearly equivalent to that consumed for generating electric power. Nations with large coal deposits have the option of coal conversion to meet their transportable fuel demands. But these processes themselves consume huge amounts of energy and produce undesirable combustion by-products. Therefore, this represents a major opportunity to apply nuclear energy for both the environmental and energy conservation reasons. Because the most desirable coal conversion processes take place at 800 degree C or higher, only the High Temperature Gas-Cooled Reactors (HTGRs) have the potential to be adapted to coal conversion processes. This report provides a discussion of this utilization of HTGR reactors

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  12. Control of the geomorphology and gas hydrate extent on widespread gas emissions offshore Romania (Black Sea)

    Science.gov (United States)

    Riboulot, V.; Cattaneo, A.; Sultan, N.; Ker, S.; Scalabrin, C.; Gaillot, A.; Jouet, G.; Marsset, B.; Thomas, Y.; Ballas, G.; Marsset, T.; Garziglia, S.; Ruffine, L.; Boulart, C.

    2016-12-01

    The Romanian sector of the Black Sea deserves attention because the Danube deep-sea fan is one of the largest sediment depositional systems worldwide and is considered the world's most isolated sea, the largest anoxic water body on the planet and a unique energy-rich sea. Due to the high sediment accumulation rate, presence of organic matter and anoxic conditions, the Black sea sediment offshore the Danube delta is rich in gas and thus show BSR. The cartography of the BSR over the last 20 years, exhibits its widespread occurrence, indicative of extensive development of hydrate accumulations and a huge gas hydrate potential. By combining old and new datasets acquired in 2015 during the GHASS expedition, we performed a geomorphological analysis of the continental slope north-east of the Danube canyon that reveals the presence of several landslides inside and outside several canyons incising the seafloor. It is a complex study area presenting sedimentary processes such as seafloor erosion and instability, mass wasting, formation of gas hydrates, fluid migration, gas escape, where the imprint of geomorphology seems to dictate the location where gas seep occurs. . Some 1409 gas seeps within the water column acoustic records are observed between 200 m and 800 m water depth. No gas flares were detected in deeper areas where gas hydrates are stable. Overall, 93% of the all gas seeps observed are above geomorphological structures. 78% are right above escarpment induced by sedimentary destabilizations inside or outside canyons. The results suggest a geomorphological control of degassing at the seafloor and gas seeps are thus constrained by the gas hydrates stability zone. The stability of the gas hydrates is dependent on the salinity gradient through the sedimentary column and thus on the Black Sea recent geological history. The extent and the dynamics of gas hydrates have a probable impact on the sedimentary destabilization observed at the seafloor.

  13. Flow dynamics around downwelling submarine canyons

    Directory of Open Access Journals (Sweden)

    J. M. Spurgin

    2014-10-01

    Full Text Available Flow dynamics around a downwelling submarine canyon were analysed with the Massachusetts Institute of Technology general circulation model. Blanes Canyon (northwestern Mediterranean was used for topographic and initial forcing conditions. Fourteen scenarios were modelled with varying forcing conditions. Rossby and Burger numbers were used to determine the significance of Coriolis acceleration and stratification (respectively and their impacts on flow dynamics. A new non-dimensional parameter (χ was introduced to determine the significance of vertical variations in stratification. Some simulations do see brief periods of upwards displacement of water during the 10-day model period; however, the presence of the submarine canyon is found to enhance downwards advection of density in all model scenarios. High Burger numbers lead to negative vorticity and a trapped anticyclonic eddy within the canyon, as well as an increased density anomaly. Low Burger numbers lead to positive vorticity, cyclonic circulation, and weaker density anomalies. Vertical variations in stratification affect zonal jet placement. Under the same forcing conditions, the zonal jet is pushed offshore in more uniformly stratified domains. The offshore jet location generates upwards density advection away from the canyon, while onshore jets generate downwards density advection everywhere within the model domain. Increasing Rossby values across the canyon axis, as well as decreasing Burger values, increase negative vertical flux at shelf break depth (150 m. Increasing Rossby numbers lead to stronger downwards advection of a passive tracer (nitrate, as well as stronger vorticity within the canyon. Results from previous studies are explained within this new dynamic framework.

  14. Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Hang; Wang Jie; Laurien, E.

    2010-01-01

    The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used